ML20093N236
| ML20093N236 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 10/09/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Vermont Yankee |
| Shared Package | |
| ML20093C502 | List: |
| References | |
| DPR-28-A-083 NUDOCS 8410310577 | |
| Download: ML20093N236 (86) | |
Text
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Z UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555 j
c VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 83 License No. DPR-28 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Vermont Yankee Nuclear Power Corporation (the licensee) dated January 23, 1984 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-28 is hereby amended to read as follows:
8410310577 941009 PDR ADOCK 05000271 P
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 83, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of April 1,1985.
FOR THE NUCLEAR REGULATORY COMMISSION
~ ~.3 ' ' '.,_ -
Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October R,1984 1
i e
4 k
ATTACHMENT TO LICENSE AMENDMENT NO. 83 FACILITY OPERATING LICENSE f;0. DPR-28 DOCKET NO. 50-271 i
Revise the Techn'ical Specifications as.follows:
Remove Insert iii iii iv iv i
y 2
2 2a 4a 4b 46 46 58 58 65 65 147 through 160 147 through 160 160a through 1601 161 161 161a 162 through 172 162 through 172 1
172a through 1721 180c 180c i
180h 180h 188 188 197 197 1
198 198 198a 198a 200 200 201 201 206 206 207a 207a 208 through 215 208 through 222 1
d
e VYNPS TABLE OF 00NTENTS (continued) 1 LIMITING CONDITIONS OF OPERATION Page No.
Surveillance l
F.
Automatic Depressurization System.........................
92 F
G.
Reactor Core Isolation Cooling System (RCIC)..............
93 G
H.
Minimum Core and Containment Cooling System Availability..
94 H
I.
Maintenance of Filled Disc harge F1pe......................
95 I
t r
3.6 REACTOR COOLANT SYSTEM........................................
105 4.6 A.
Pressure and Temperature Limitations......................
105 A
5.
Coo l a n t C hem i st ry.........................................
106 B
C.
Coolant Leakage...........................................
1 08 C
D.
S a f e t y a nd R e l i ef Va 1 ve s..................................
108 D
E.
Structural Integrity......................................
109 E
F,.
Jet Pumps.................................................
109 F
G' Single-Loop Operation.....................................
110 H.
Rec i rcula t io n 5 y st em......................................
110 I.
S hoc k S u r p a c a so r s.........................................
110s I
J.
T he rmal-Hyd rau lic S tab ( 11 ty...............................
110b J
3.7 STATION ColiTAINNENT SYSTEMS..............'....................
126 4.7 A.
Priarcy Containment.......................................
126 A
B.
d t a nd by Ga s T re a t me n t.....................................
1 30 E
C.
S econda ry Co ntai nme nt Sy st em..............................
131 C
D.
Primary Containment Isolation Va1ves......................
132 D
L 4
Acanitoent No. hr 83
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a VYMPS TABLE OF CONTENTS (continued) 1 i
LIMITING CONDITIONS OF OPERATION Page No.
Surveillance l
3.8 RAD IOA CTIVE E F F LUE NTS.........................................
147 4.8 A.
Liquid Effluents: Co nc e nt ra t i o n..........................
147 A
5.
Liquid Effluents:
Dose...................................
148 B
C.
Liquid Rad aste Treatment.................................
149 C
D.
Liquid Holdup Tanks.......................................
149 D
E.
Gaseous Effluents: Dose Rate.............................
150 E
F.
Geseous Effluents: Dose From Noble Gases.................
151 F
G.
Gaseous Effluents: Dose From Iodine-131. Iodine-133 Tritium, and Radioactive Materials in Particulate Fora....
152 G
H.
Caseous Radwaste Treatment................................
152 H
I.
Ventilation Exhaust Treatment.............................
153 1
J.
Explosive Gas Mixture.....................................
153 J
K.
Steam Jet Air Ejector (SJAE)..............................
153 K
L.
Primary Containment.....<.................................
154 L
M.
Total Dose................................................
155 M
N.
Solid Radioactive Waste...........................
156 N
l 3.9 RADIOACTIVE EFFLUENT HONITORING SYSTEMS.......................
161 4.9 1
l l
A.
Liquid Effluent Instrumentation...........................
161 A
B.
Ca seous E f fluent Inst rume ntat ion..........................
161a B
C.
Radiological Envi ronme ntal Moni toring Program.............
161a C
D.
La nd U se Ce n s u s...........................................
162 D
E.
. I n t e rcomp a ri so n P rog ram...........................'. ~.......
163 E
1 Anandment No. d), jf, 83
-iv-i
t f
VYNPS TABLE OF CONTENTS (continued)
LIMITING CONDITIONS OF OPERATION Page No.
Surveillance 3.10 AUXILIARY ELECTRICAL POWER SYSTEMS............................
173 4.10 A.
No rma l Ope ra t i o n..........................................
173 Al B.
Ope ration with Inope ra blo Compone nt s......................
176 B
C.
Diesel Fue1...............................................
177 C
i 3.11 REACTOR FUEL ASSEMBLIES.......................................
180s-4.11 A.
Average Planar LHCR.......................................
-180s A-i B.
LHGE......................................................
180b B
C.
MCPR......................................................
180c C
6 3.12 REFUELING AND SPENT FUEL KANDLING.............................
181 4.12 3
A.
Re f ue li ng I n t e r1oc ks......................................
181 A
B.
Co re Mo ni t o ri ng...........................................
182 B
C.
Fuel S to rage Pool Wate r Leve1.............................
183 C
D.
Cont rol Rod and Cont rol Rod Drive Mai ntena nce.............
184 D
E.
Ex t e nd ed Co re Mai nt e na nc e......... ;.......................
184 R
F.
Fuel Hovement.............................................
185 F
G.
C ra ne O pe ra b i l i t y.........................................
185 G
j H.
S pent Fuel Fool Wa ter Tempe ra tu re.........................
185s H
1 1
3.13 FIRE PROTECTION SYSTEM........................................
187b 4.13 5.0 DESIGN FEATURES....................................................
188 6.0 ADMINISTRATIVE CONTR0LS............................................
190 f
6 Amendment No.
o3 v-
)
VYNPS G.
Inst rument Functional Test - An instrument K.
Operable - A system, subsystem, train, component functional test shall bet or device shall be operable or have operability when it is capable of performing its specified j
1.
Analog channels - the injection of a signal function (s). Implicit in this definition shall be into the channel as close to the sensor as the assumption that all necessary attendant
., practicable to verify operability including instrumentation, controle, normal and emergency alarm and/or trip functions.
electrical power sources, cooling or seal water.
l lubrication or other auxiliary equipment that are
- 2. 1 Bistable channels - the injection of a signal required for the system, subsystes, train.
l into the sensor to verify operab111ty component or device to perform its function (s) are including alarm and/or trip functions, also capable of performing their related support function (s).
H.
Loalc Systen Functional Test - A logic systen functional test means a test of all relays and L.
Operatina - Operating means that a system or 1
contacts of a logic circuit from sensor to component is performing its intended functions in activated device to insure all components are its required manner.
operable per design intent. Where possible, action will go to completion, i.e., pumps will be M.
Operatina Cycle - Interval between the end of one started and valves opened.
refueling outage and the end of the next subsequent refueling outage.
I.
Minimum Critical Power Ratio - The minimue l
critical power ratio is defined as the ratio of M.
Primary Containment Integrity - Primary that power in a fuel assembly which is calculated containment integrity means that the drywell and to cause some point in that assembly to experience pressure suppression chamber are intact and all of boiling transition as calculated by application of the following conditions are satisfied the CEXL correlation to the actual sesembly operating power (Reference NEDO-10958).
1.
All manual containment isolation valves on lines connecting to the reactor coolant J.
Mode - The reactor mode is that which is system or containment which are not required established by the mode selector switch.
to be open during accident conditions are i
closed.
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,l Amendment No. 61, 3 6, 00 l.
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VTWFS 2.
At least one door in each airlock is closed and sealed.
3.
All automatic containment isolation valves are operable or deactivated in the isolated
.msition.
4.
All blind flanges and manways are closed.
O.
Protective Instrumentation Definitions 1.
Inst rument Channel - An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transsit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.
2.
Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective
{
t rip function. A trip system may requirp one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action.
l Initiation of protective action may require j
the tripping of a single trip system or the coincident tripping of two trip systems.
i Aasadment No.
83 2a t
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VYNPS BB.
Storce Check - The qualitative assessment of GG.
Radioactive Materisi - Any material or combination channel response when the channel sensor is of materials which spontaneously emits ionizing exposed to a radioactive source.
radiation and in which the specific activity is greater than 0.002 microcuries/ gram of material or CC.
Dese Equivalent I-131 - The dose equivalent I-131 any material in which the total estimated activity chall be that concestration of I-131 is greater than 5 microcuries is classified as a (cierocurie/ gram) which alone would produce the radioactive material.
crea thyroid dose as the quantity and isotopic cixture of I-131, 1-132, 1-133, I-134 and I-135 HH.
Contamination cetually present. The thyroid dose conversion i
fcctors used for this calculation shall be those 1.
Removable radioactive contamination shall be listed in NRC Regulatory Guide 1.109, Rev. 1, considered significant and unreleasable f rom October 1977.
the owner controlled area if the level, when averaged over 300 square centimeters, exceeds
,DD.
Solidification - Solidification shall be the 220 dpa/ square centimeter for beta gamma, conversion of wet wastes into a form that meets and 22 dpm/ square centimeter for alpha chipping and burial ground requirements. Suitable emitting radionuclides.
forms include dewatered resins and filter sludges.
2.
Fixed contamination shall be considered
.EE.
Mraber(s) of the Public - Members of the public significant and unreleasable f rom the owner chall include all persons who are not controlled area if the dose rate at any occupationally associated with the plant. This accessible surface exceeds 0.5 area / hour.
category does not include employees of the utility, its contractors or vendors. Also II.
Of f-Site Dose Calculation Manual (ODCM) - A manual eluded from this category are casual visitors to containing the current methodology and parameters ue plant and persons who enter the site to used in the calculation of off site doses due to 6 arvice equipment or to make deliveries.
radioactive gaseous and liquid effluents, in the calculation of gaseous and 11guld effluent FF.
Site Boundary - The site boundary is shown in monitortug alare/ trip setpoints, and in the Figure 2.2-5 in the FSAR.
conduction of the envf.conmental radiological monitoring program.
i Amendesnt No.
83 4a l
s
V1fNPS JJ.
Process Control Program (PCF) - A process control I.L.
Ventilation Exhaust Treatment System - The l
program shall contain the sampling, analysis, Radwaste Building and A0G Building ventilation tests, and determinations by which wet radioactive llEPA filters are ventilation exhaust treatment waste from liquid systems is assured to be systems which have been designed and installed to converted to a form suitable for off site disposal.
reduce radioactive material in particulate form in gaseous effluents by passing ventilation air KK.
Gaseous Radwaste Treatment System - The Augmented through HEPA filtera for the purpose of removing j.
Off-Gas System (A0G) is the gaseous radwaste radioactive particulates from the gaseous exhaust treatment system which has been designed and stream prior to release to the environment.
Installed to reduce radioactive gaseous effluents Engineered safety feature atmospheric cleanup T
by collecting primary coolant system off gases systems, such as the Standby Gas Treatment (S5GT) from the primary system and providing for delay or Syates, are not considered to be ventilation holdup for the purpose of reducing the total exhaust treatment system components, radioactivity prior to release to the environment.
NN.
Vent / Purging -. Vent / Purging is the controlled process of discharging air or gas f rom the primary containment to control temperature, pressure, humidity, concentration or other operating conditions.
0 0
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Amndment No.
83 4b 1
4
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VYNPS TABLE 3.2.4 OFF-GAS SYSTEM ISOLATION INSTRUNENTATION l
Minimum Number of Required Action When Oparable Instrument '
Minimum Condition for Chanrels per Trip System Trip Function Trip Settina Operation Are Not Met 1
Time Delay (Stack Off-Gas Valve 1 2 minutes Note 1 Isolation) (15TD & 16TD) 1 30 minutes Note 1 1
Trip System Logic At least one of the radiation monitors between the charcoal bed system and the plant stack shall be Nato 1 operable during operation of the augmented of f gas system.
If this condition cannot be met, continued operation of the augmented of f gas system is permissible for a period of up to 7 days provided that at least one of the stack monitoring systems is operable and of f gas aystem temperature and pressure are measured continuously.
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The APRM rod block trip is flow referenced and prevents a significart reduction in MCPR especially during operation at reduced flow. The APRM provides gross core protection; i.e.,
limits the gross core power increase f rom withdrawal of 4j control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the fuel cledding integrity safety limit.
The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip catting is less than a factor of 10 above the indicated level. Analysis of the worst case accident results in rod block cction before MCPR approaches the fuel cladding integrity safety limit.
A downscale indication on an APRM or IRN is an indication the instrument has failed or the instrument is not sensitive cnorgh.
In either case the instrument will not respond to changes in control rod motion and thus control rod motion lo provented.
To provent excessive clad temperatures for the small pipe break, the HPCI or Automatic Depressurisation System must functica since for these breaks, reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The arrangement of the tripping contacts is such as to provide this function when necessary and cinicite spurious operation. The trip settings given in the specification are adequate to assure the above criteria cra met.
The specification preserves the ef fectiveness of the systes during periods of maintenance', testing, or i calibration and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.
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l Amendment No. Jr.
83 65 i
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VYNPS 3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS 3.8 RADI0 ACTIVE EFFLUENTS 4.8 RADIOACTIVE EFFLUENTS r
Applicability Applicability j
1 Applies to the release of all radioactive Applies to the required surveillance of all effluents from the plant.
radioactive effluents released from the plant.
Objective Objective To assure that radioactive effluents are kept "as To ascertain that all radioactive effluents low as is reasonably achievable" in accordance released f roe' the plant are kept "as low as is with 10CFR50, Appendix I and, in any event, are reasonably achievable" in accordance with 10CFR50, within the limits specified in 10CFR20.
Appendix 1 and, in any event, are within the limits specified in 10CFR20.
Specification Specification A,
Liquid Effluents: Concent ration A.
Ligeld Effluents: Concent ration 1.
The concentration of radioactive 1.
Radioactive material in liquid waste material in liquid ef fluents released shall be sampled and analysed in from the site shall be limited to the accordance with requirements of Table concentrations specified in 10CFR Part 4.8.1.
The results of the analyses 20, Appendix 5. Table II, Column 2 for shall be used in accordance with the radionuclides other than noble gases and methods in the ODCN to assure that the 2x10~4 uC1/m1 total activity concentrations at the point of release concentration for all dissolved or are limited to the values in entrained noble gases.
Specification 3.8.A.1.
Amendrent No.
83 147
l VYNPS e
3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS 2.
With the concentratica of radioactive material in liquid effluents released f rom the site exceeding the limits of Specification 3.8.A.1, immediately take i
action to decrease the release rate of radioactive materials and/or increase the dilution flow rate to restore the concentration to within the above limits.
B.
Liquid Effluents: Dose B.
Liquid Effluents: Dose 1.
The dose or dose commitment to a member 1.
Cumulative dose contributions shall be of the public from radioactive materials determined is accordance with the in 11guld effluents released from the methods in the ODCN at least once per site shall be limited to the following:
month if releases during the period have occurred.
a.
During any calendar quarter:
less than or equal to 1.5 ares to the total body, and less than or equal to 5 ares tv any organ, and b.
During any calendar year i
)
l less than or equal to 3 area to the total body, and less than or equal to 10 meen to any organ.
Amendaant No.
83 148
^
I VYNPS
/
3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS
~
C.
Liquid Radwaste Treatment C.
Liquid Radweste Treatment 1.
The liquid radvaste treatment system 1.
See Specification 4.8.8.1.
shall be used in its designed modes of operation to reduce the radioactive materials in the liquid waste prior to its discharge when the estimated doses due to the liquid effluent from the site, when averaged with all other liquid release over the last month, would exceed 0.06 area to the total body, or 0.2 area to any organ.
D.
Liquid Holdup Tanks D.
Liquid Holdup Tanks 1.
The quantity of radioactive material 1.
The quantity of redioactive material contained in any outside tank
- shall be contained in each of the liquid holdup limited to less than or equal to 10 tanks
- shall be determined to be within curies, excluding tritium and dissolved the Ilmits of Specification 3.8.D.1 by or entrained noble gases.
analyzing a representative sample of the tank's content at least once per week or when radioactive materials are being added to the tank.
(NOTE: Tanks ir.cluded in this Specification are only those outdoor tanks that are not surrounded by liners, dikss, or walls capable of holding the tank's contents, or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.
Amendment No.
83 149
t s
VYNPS 3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEII. LANCE REQUIREMENTS t-2.
With the quantity of radioactive material in any outside tank
- exceeding j
the limit of Specification 3.8.D.1, j
immediately take action to suspend all additions of radioactive material to the tank. Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, reduce the tank contents to within the limit.
E.
Gaseous Effluents: Dose Rate E.
Gaseous Effluents: bose Rate 1.
The dose rate due to radioactive 1.
The. dose rate due to noble gases in materials released in gaseous effluents gaseous affluents shall be determined to be within the limits of Specification f rom the site to areas at and beyond the 3.8.E.1 in accordance with the methods site boundary shall be limited to the followings in the ODCH.
a.
For noble gases; less than or equal 2.
The dose rate due to Iodine-131, io 500 ares /yr to the total body Iodine-133, tritium and radionuclides in and less than or equal to 3,000 particulate form with half-lives greater ares /yr to the skin, and than 8 days in gaseous effluents shall be determined to be within the limits of b.
For Iodine-131. Iodine-133, t ritium Specification 3.8.E.1 in accordance with and radionuclides in particulate the methods in the ODCM by obtaining form with half-lives greater than 8 representative samples and performing days; less than or equal to 1,500 analyses in accordance with the sampling ares /yr to any organ.
and analysis program specified in Table i
4.8.2.
2.
With the dose rate (s) exceeding the j
above limits, immediately taka actice to I
decrease the release rate to within the 11mits of Specification 3.8.E.1.
I i
i Acandment No.
83 150 e
I
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VYNPS i
i L
3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS f
F.
Gaseous Effluents: Dose from Noble Gases F.
Caseous Effluents: Dose from Noble Cases 1.
The air dose due to noble gases released 1.
Cumulative dose contributions for the in gaseous effluents from the site to total time period shall be determined la areas at and beyond the site boundary accordance with the methods in the ODCN shall be limited to the followings at least once every month.
s.
During any calendar quarters i
I less than or equoi to 5 stad for gamma radiation, and less than or equal to 10 mrad for i
beta radiation, and b.
During any calendar year:
less than or equal to 10 mrad for gamma radiation, and t
less than or equal to 20 mrad for beta radiation.
W I
Amendment No.
83 151 1
~
VYNPS g-3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIRENENTS C.
Caseous Effluents: Dose from Iodine-131, C.
Caseous Effluents: Dose f rom Iodine-131, lodine-133, Tri tium, and Radionuclides in lodine-133, Tritium, and Radionuclides in Particulate Form Particulate Form 1.
The dose to a member of the public f rom 1.
Cumulative dose contributions for the Iodine-131 Iodine-133, tritium, and total time period shall be determined in radionuclides in particulate form with accordance with the methods in the ODCM half-lives greater than 8 days in at least once every month.
gaseous effluents released from the site to areas at and beyond the site boundary shall be limited to the following:
s.
During any calendar quarter:
less than or equal to 7.5 aren to any organ, and b.
During any calendar year:
less than or equal to 15 area to any organ.
H.
Caseous Radwaste' Treatment H.
Caseous Radwaste Treatment 1.
The Augmented Off-Cas System (AOC) shall 1.
The readings of the relevant inst rument be used in its designed mode of shall be checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the operation to reduce noble gases in main condenser SJAE is in use to ensure gaseous vaste prior to their discharge that the AOC is functioning.
whenever the main condenser steam jet air ejector (SJAE) is in operation.
i.
Amendment No.
83 152
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3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE EEQUIREMENTS
[
+:
I.
Ventilation Exhaust Treatment I.
Ventilation Exhaust Treatment 1.
The AOG and Radwaste guilding 1.
See Specification 4.8.F.1 for Ventitation Filter (IE!PA) Systems shall surveillance related to AOC rad Radwaste be used to reduce particulate materials Building ventilation filter system 1
in gaseous waste prior to their ope ration.
discharge from those buildings when the l
estimated doses due to gaseous effluent releases f rom the site to areas at and
]
beyond the site boundary would exceed O.3 ares to any organ over one month.
J.
Explosive Gas Mixture J.
Explosive Gas Mixture i
1.
If the hydrogen concentration in the 1.
The concentration of hydrogen in the j
of f gas downstream of the operating off gas system downstress of the recombiner reaches four percent, take recombiners shall be continuously I
appropriate action that will restore the monitored by the hydrogen monitor f
concentration to within the limit within required operable by Table 3.9.2.
48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
{
K.
Steen Jet Air Ejector (SJAE)
K.
Stese Jet Air Elector (SJAE) 1.
Gross radioactivity release rate from 1.
The gross radioactivity release rate j
the SJAE shall be limited to less than shall be continuously monitored in or equal to 0.16 C1/sec (af ter 30 accordance with Specification 3.9.5.
j minutes decay).
2.
The gross radioactivity release rate of 2.
With the gross radioactivity release noble gases from-the SJAE shall be rate at the SJAE exceeding the above determined to be within the limit of l
limit, restore the gross radioactivity Specification 3.8.K.1 at the following i
release rate to within its limit within frequencies by performing an isotopic j
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least Hot Standby analysis (for Xe-138, Xe-135, Xe-133, within the subsequent 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Kr-88, Kr-85e, Kr-87) on a representative sample of gases taken at the discharge.
Assndesnt No.
83 153 9
a t
~.
3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS 3.
With the gross radioactivity release rate at the SJAE greater than or equal to 1.5 C1/sec (af ter 30-minute decay).
restore the gross radioac*1vity release rate to less than 1.5 C1/sec (after 30 minute decay), or be in Hot Standby within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following an increase of 25% or 5000 microcuries/sec, whichever is greater, in steady-state activity levels during steady state reactor operation, as indicated by the SJAE monitor.
L.
1.
If the primary containment is to be 1.
The primary containment shall be sampled Vented / Purged, it shall be Vented / Purged prior to venting / purging per Table through the Standby Cas Treatment System 4.8.2, and if the results indicate whenever the airborne radioactivity radioactivity levels in excess of the levels in containment exceed the levels limits of Specification 3.8.L.1, the specified in 10CFR20, Appendix B. Table containment shall be aligned for I, Column 1 and notes 1-5 thereto.
venting / purging through the Standby Cas Treatment System. No sampling shall be 2.
With the requirements of Specification required if the venting / purging is 3.8.L.1 not estisfied, immediately through the Standby Gas Treatment (SBCT) suspend all Venting / Purging of the System.
containment.
3..
During normal refueling and maintenance outages when primary containment is no longer required, then Specification 3.8.C shall supersede Specifications 3.8.L.1 and 2.
- Amendment No.
83 154 t
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3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS 4
M.
Total Dose M.
Total Dose 1.
The dose or dose commitment to a member 1.
Cumulative dose contributions f rom
)
of the public* from all station sources liquid and gaseous effluents shall be
{
is limited to less than or equal to 25 determined in accordance with mres to the total body or any organ Specifications 4.8.B.1, 4.8.F.1, and I
(except the thyroid, which is limited to 4.8.G.I.
i less than or equal to 75 aren) over a calendar year.
2.
Cumulative dose contributions from direct radiation f rom plant sources 2.
With the calculated dose f rom the shall be determined in accordance with release of radioactive materials in the' methods in the 00CM. This liquid or gaseous effluents exceeding requirement is applicable only under twice the limits of Specifications conditions set forth in Specification 3.8.B.1.a. 3.8.B.1.b, 3.8.F.1.a.
3.8.M 2.
- 3. 8.F.1. b. 3. 8.G.1.s. or 3. 8.G.1. b, I
calculations should be made, including direct radiation contributions from the station to determine whether the above limits of Specification 3.8.M.1 have been exceeded.
s i
i CNOTE: For this Specification a member of the public may be taken as a real individual accounting for his actual cctivitt's.
e Amendnent No.
OO 155 l
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3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS N.
Solid Radioactive Waste M.
Solid Radioactive Waste i
1.
The solid radwaste system shall be used 1.
Verification of solidification of wet in accordance with a Process Control waste shall be performed as required and Program as described in Section 6.12 to in accordance with the Process Control j
process wet radioactive vaste (opent Program.
l resins / filter sludges) to meet shipping i
and burial ground requirements.
2.
With the provisions of Specification 3.8 N.1 not satisfied, suspend shipments of defectively processed or defectively packaged solidified wet radioactive wastes from the site.
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Amendment No.
83 156 i
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VYNPS TABLE 4.8.1 RADIDACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM l
l Lower Limit Minism of Detection i
Sampling Analysis Type of Activity (LIA)
I Liquid Release Type Frequency Freque ncy Analysis (uC1/el)s Estch Waste Release Tankab Prior to Prior to Principal Gamma 5 x 10-7 each release each release Emittered I
Each Batch Each Estch 1-131 -
1 x 10-6 One Batch once per Dissolved and 1 x 10-5 per month month Entrained Gases sampled (Goana Enttters) prior to a relea se i
Prior to once per H-3 1 x 10-5 each release sonth c
Each Batch Composite Cross Alpha 1 x 10~7 i
Prior to once per Sr-89 Sr-90 5.x 10-8 1
each release quarter j
Each Batch Compositec Pe-55 1 x 10-0' I
I i
Amendment No.
83 157 I
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i i
VYNFS TABLE 4.8.1 (continued)
TABLE NOTATION o.
The LLD is the smallest concentration of radioactive material in a sample that will yield a not count, above cystem background, that will be detected with 951 probability with only 51 probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (whien may include radiochemical separation):
4.66
- S b LLD =
E*V*K*Y*e where:
LLD = the lower limit of detection as defined above (microcuries or picocuries/ unit mass or volume) b
= the standard deviation of the background counting rate or of the counting rate of a blank sample as S
appropriate (counts / minute)
E
= the counting ef ficiency (counts / disintegration)
V
= the sample size (units of mass or' volume)
K
= 2.22 x 106 disintegrations / minute / microcurie or 2.22 disintegrations / minute / picocurie as applicable Y
= the f ractional radiochemical yield (when applicable) g A
= the radioactive decay constant for the particular radionuclide (/ minute)
At
= the elapsed time between sample collection and analysis (minutes)
Amendment No.
83 158 I
4 4
VYNFS TABLE 4.3.1 (continued)
TABLE MDTATION Typical values of E V, Y anddt can be used in the calculation. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples.
I Analysis shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.
g j
Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or i
other uncontrollable circumstances may render these LLDs unavailable.
i It should be recognised that the LLD is defined as a "before the f act" Ifnit representing the capability of a measurement system and not as an "af ter the fact" limit for a particular measurement. This does not preclude the calculation of an "af ter the fact" LLD for a particular measurement based upon the actual parameters for the sample in question and appropriate decay correction parameters such as decay while sampling and during analysis.
b.
A batch release is the discharge of 11guld wastes of a discrete volume. Prior to sampling for analysis, each batch shall be isolated and then thoroughly atxed to assure representative sampling.
A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid c.
waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release, d.
The principal samma eettters for which the LLD specification will apply are exclusively the following I
radionuclides:
Mn-54, Fe-59, Co-58, Co-60, Zn-65 No-99, Cs-134, Cs-137 Ce-141, and Ce-144. This list does not mean that only these nuclides ar'e to be detected and reported. Other peaks which are esasurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level, but as "not detected". When unusual circumstances result in LLDs higher than required, the reasons shall be docunented in the Semiannual Effluent Release Report.
Amendment No.
83 159 j
i
VYNPS TABLE 4.8.2 RADI0 ACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Ministas of Detection Sampling Analysis Type of Activity (LLD)
Casecun Release Type Frequency Frequency Analysis (uC1/al)a_
e A.
Secan Jet Air Once per Once per Xe-138, Xe-135, 1 x 10-4 Ejector week week Xa-133. Kr-88, Crab Sample Kr-87, Kr-85N B.
Costairement Purge Prior to Prior to Principal Gaassa 1 x 10-4 each release each release Emitterad Each Purge Each Purge Grab Sample C.
Main Plant Stack Once per Once pe:-
Principal Caerna 1 x 10-4 nonthe monthe Emittered
/
Grab Sample H-3 1 x 10-6 Continuous
- Once per I-131f 1 x 10-12 weekb Charcoal Sample Continuous
- Once per weekb Principal Camma 1 x 10-11 Particulate Emittered Sample (I-131)
Continuous
- Once per Cross Alpha 1 x 10-11 month Composite Particulate Sample Amend:ent No.
83 160
VYNPS TABLE 4.8.2 (ccatinued)
RADI3 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAN Lower Limit Minimum of Detection Sampling Analysis Type of Activity (LLD) e Giseous Release Type Frequency Frequency Analysis (uC1/al)a_
C.
(continued)
Continuous
- Once per Sr-89, Sr-90 1 x 10-11 quarter Composite Particulate Sample Continuous Noble Gas Noble Cases 1 x 10-5 Monitor Gross Bets or Gamma 00 Amendment No.
160s i
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VYNpS TABLs 4.8.2 (continued)
TABLE NOTATION o.
See footnote a. of Table 4.8.1.
b.
Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> af ter removal from samplers. Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup or thetual power change exceeding 25% of rated thermal power in one hour, and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing the samples. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10.
This requirement to sample at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for 7 days applies only if: (1) analysis shows that the dose equivalent 1-131 concentration in the primary coolant has increased more than a factor of 3 and the resultant concentration is at least 1 x 10~1 uC1/al; and (2) the noble gas monitor shows that effluent activity has increased more than a factor of 3.
c.
Sampling and analyses shall also be performed following shutdown, startup, or a thermal power change exceeding 25% of rated thermal power per hour unless: (1) analysis shows that the dose equivalent 1-131 concentration in the primary coolant has not increased more than a factor of 3 and the resultant concentration is at least 1x 10-1 uC1/al; and (2) the noble gas monito:c shows that effluent activity has not increased more than a factor of 3.
d.
The principal samma emitters for which the LLD specification will apply are exclusively the following radionuclides:
Kr-87, Kr-88, Xe-133, Ie-133s, Xe-135 and Ie-138 for gaseous emissions, and Nn-54, Fe-59, Co-58, Co-60, Zn-65, No-99, Cs-134, co-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below LLD for the i
analysce should not be reported as being present at the LLD level for that nuclide, but as "not detected". When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Sealannual Effluent Release Report.
s.
The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.8.E.1, 3.8.F.1 and 3.8.G.I.
5 Amendment No.
83 160b i
4 L
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VYllPS TABLE 4.8.2 (continued)
TABLE NOTATION f.
The gaseous vaste sampling and analysis program does not explicitly require sampling and analysis at a specified i
LLD to determine the 1-133 release. Estimates of 1-133 releases shall be determined by counting the weekly l
charcoal sample for 1-133 (as well as 1-131) and assume a constant release rate for the release period.
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Amendment No.
83 160c l
VYWpS
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B&SES:
3.8 RADIDACTIVE EFFLUENTS A.
Liquid affluents: Concent rat ion This specification is provided to ensure that the concentration of radioactive materials released in liquid weste ef fluents from the site above background (at the point of discharge f rom the plant discharge into Connecticut River) will be less than the concentration levels specified in 10CFR Part 20. Appendix B. Table II, Column 2.
This 11altation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within (1) the Section II.A design objectives of Appendix 1,10CFR Part 50, to a member of the public, and (2) the limits of 10CFR Part 20.106 (e) to the population.
The concentration limit for noble gases is based upon the assunption that Ze-135 is the controlling radionuclide and its NFC in air (submersion) was converted to an equivalent concentration in water uslag the International Commission on Radiological Protection (ICEF) Publication 2.
B.
Liquid Effluents: Dose This specification is provided to implement the requirements of Sections II.A. III.A, and IV.A of Appendix I, 10CFR Part 50.
The Limiting Condition for Operation implements the guides set forth in Section II. A of Appendix 1.
The requirements provide operating flexibility and at the same time laplement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid of fluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix 1, i.e.,
that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. In addition.
00 Aacudsent No.
160d t
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VYMPS 308 (Continued) i j
there is reasonable assurance that the operation of the facility will not result la radionuclide i
concentrations in potable drinking water that are la excess of the requirements of 40CFR 141. No drinklag water supplies drawn f rom the Connecticut River below the plant have been identified. The appropriate dose J
equations for implementation through requirements of the Specification are described la the Vermont Yankee Of f-Site Dose Calculation Manual. The equations sPecified la the 00CN for calculating the doses due to the actual release rates of radioactive materials in liquid effluents were developed f rom the methodology i
provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man f rom Routine Releases of Reactor
]
Rf fluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix 1". Revision 1, October l
1977 and Regulatory Guide 1.113, "Estimatt::g Aquatic Lispersica af Effluents from Accidental sad koutine Reactor Releases for the Purpose of Impleenenting Appendix I", Revision 1. April 1977.
C.
Liquid Radusste Treatment i
The requirement that the appropriate portions of this system as indicated in the ODCN be used, when 4
l specified, provides assurance that the releases of radioactive materials in 11guld effluents will be kept
'as low as is reasonably. achievable". This specification implements the requirements of 10CFR Part 50.36a
. and the design objective given in Section II.D of Appendix I to 10CFR Part 50. The specified limits i
governing the use of appropriate portions of the 11guld redunste treatment system were'opecified as a suitable f raction of the dose design objectives set forth in Section II.A of Appendix I,10CFR Fort 50, for l
11guld effluents.
O.
Lieuid Moldup Tanks i
The tanks listed in this Specification include all outdoor tanks that contata radioactivity that are not surrounded by liners, dikes, or walls capable of holding the tank contents, or that do not have tank j
overflows and surrounding ares drains connected to the 11guld radwaste treatment system.
Restricting the quantity of radioactive material contained in the specified tanks provides assurance that l
In the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less j
than the limits of 10CFR Part 20, Appendix 5. Table II, Column 2, at the nearest portable water, supply and l
the nearest surface water supply in an unrestricted area.
l l
t Amendment No.
83 160s I
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7 VYNPS
~
e 3.8 (Continued) e E.
Gaseous Effluents: Dose Rate This specifiestion is provided to ensure that the dose at any time at and beyond the site boundary from gaseous ef fluents will be within the annual dose limits of 10CFR Part 20.
The annual dose limits are the doses associated with the concentrations of -10CFR Part 20 Appendix 3 Table II, Column 1.
These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of member (s) of the public either within or outside the site boundary, to annual i
average concentrations exceeding the limits specified in Appendix 3. Table II of 10CFR Fort 20 [10CFR Part 20.106(b)]. For member (s) of the public who may at times be within the site boundary, the occupancy of the individual wLil be sufficiently low to compensate for any increase in the atsaspheric diffusion f actor above that for the site boundary. The specified limits as determined by the methodology in the ODCH, restrict, at all times, the correspcading gamma and beta dose rates above background to a member of the public at or beyond the site boundary to (500) aces / year to the total body or to (3,000) area / year to the skin.
Specification 3.8.E.b also restricts, at all times, comparable with the length of the ' sampling periods of Table 4.8.2 the corresponding thyroid dose rate above background to an infant via the cow milk-infant pathway to 1500 mrea/ year for the nearest cow to the plant.
F.
Gaseous Effluents: Dose from Noble Gases This specification is provided to implement the requirements of Sections 11.3. III.A. and IV.A of Appendix I, 10CFR Part 50.
The Limiting Condition for Operation implements the guides set forth in Section 11.3 of Appendix 1.
The requirements provide operating flexibility and at the same time leptement the guides set foeth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". The Surveillance Requiremente implement the requirement's in Section III. A of Appendix I, i.e., that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of any member of the p blic through appropriate pathways is unlikely to be substantially underestimated.
The appropriate dose equations are specified in the ODCM for calculating the doses due to the actual releases of radioactive noble gases in gaseous affluents. The 00CN also provides for determinir.g the air doses at the site boundary based upon the historical average atmospheric conditions.
Amendzent No.
83 160f
i i
VYNPS l
l 3.8 (Continued) i The equations specified in the ODCN for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents were developed f rom the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Efflutats for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I", Revision 1, October 1977 aad Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases f rom Light-k'ater Cooled Reactors," Revision 1. July 1977.
2 G.
Caseous Effluents: Dose from Iodine-131, Iodine-133 TrStium, and Radionuclides in Particulate Form This specification is provided to implement the requirements of Sections II.C. III.A. and IV. A of Appendix I, 10CFR Part 50.
The Limiting Condition for Operation are the guides set forth in Section II.C of Appendix 1.
The requirements provide operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestlasted. The equations specified in the ODCN for calculating the doses due to the actual release rates of the subject materials were also developed using the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Nan from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, j
Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases f rom Light-Wa' er Cooled t
Reactors," Revision 1, July 1977. Thes, equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine 131 Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man, in areas at and beyond its site boundary. The pathways which were examined in the development of these specifications were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent I
l consumption by man, 3) deposition onto grassy areas where silk animals and meat producing antaals graze with consumption of the allk and meat by man, and 4) deposition on the ground with subsequent, exposure of l
man.
I I
Amendment No.
83 160g i
3 4
~
-t VYNPS 3.8 (Continued)
H.
Gaseous Radwaste Treatment The requirement that the appropriate portions of the Augmented Off-Gas (A0G) System be used whenever the SJAE is in operation provides reasonable assurance that the releases of radioactive materials in gaseous ef fluents will be kept "as low as is reasonably achievable". This specification implements the i
requirements of 10CFR Part 50.36a and the design objectives of Appendix I to 10CFR Part 50.
I.
Ventilation Exhaust Treatment i
The requirement that the AOG Building and Radwaste Building HEPA filters be used when specified provides reasonable assurance that the release of radioactive materials in gaseous effluents will be.kept "as low as is reasonably achievable". This specification implements the requirements of 10CFR Part 50.36a and the design objective of Appendix 1 to 10CFR Part 50.
The requirements governing the use of the appropriate portions of the gaseous radwaste filter systems were specified by the NRC in NOR5G-0473 Revision 2 (July 1979) as a suitable f raction of the guide set forth in Sections II.B and II.C of Appendix I,10CFR Part 50, for gaseous effluents.
J.
Explosive Gas Nixture The hydrogen monitors are used to detect pessibic hydrogen buildups which could result in a possible hydrogen explosion. Automatic isolation of the offegas flow would prevent the hydrogen explosion and 4
possible damage to the augmented off gas system. Maintaining the concentration of hydrogen below its flammability limit provides assurance, that the releases of radioactive materials will be controlled.
K.
Steam Jet Air Ejector (SJAE) j Restricting the gross radioactivity release rate of gases f rom the main condenser SJAE provides reasonable j
assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specificaticn implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10CFR Part 50.
i Amendment No.
83 160h j
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1 i
8 1
VYNPS t
(
3.8 (Continued)
}..
Primary Containment (MARK I)
This specification provides reasonable assurance that releases from containment purging / venting operations will be filtered through the Standby Gas Trescuent System so that the annual dose limits of 10CFR Part 20 at the site boundary will not be exceeded. The dose objectives of Specification 3.8.G restrict i
purge / venting operations when the Standby Gas Treatment System is not in use and gives reasonable assurance' I
that all releases f rom the plant will be kept "as low as is reasonably achievable".
l M.
Total Dore j
1 l
This specification is provided to meet the dose limitations of 40CFR Part 190 that have been incorporated into 10CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Specific Report whenever the calculated doses from plant radioactive effluent's exceed twice the design objective j
doses of Appendix 1.
For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40CFR Part 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action that should result in the limitation of the annual dose to a member of the public to within the 40CFR Part 190 limits.
For the purposes of the Special Report, it may be assumed that the dose commitment to the member of the public is estimated to exceed the requirements of 40CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been coa rected), in accordance with the provisions of 40CFR Part 190.11 and 10CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed.
The variance only relates to the limits,of 40CFA Part 190, and does not-apply in any way to the other requirements for dose limitation of 10CFR Part 20, as addressed in Specification 3.8. A and 3.8.E.
An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
N.
Solid Radioactive Waste This specification implements the requirements of 10CFR Part 50.36a with respect to the hand 11.ng of solid radioactive waste (spent resin and filter sludges only). The establishment and implementation of a Process Onntrol Program (PCP), provides the operational guidelines by which proper dewatering of filter media and spent resins in preparation for of f site disposal is assured.
I Amendcant No.
00 1601 I
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l VYMPS j 3.9 LIMITING CONDITIONS FOR OPERATION 4.9 SURVEILLANCE REQUIREMENTS l
3.9 RADIO $CTIVE EFFLUENT NONITORING SYSTE!is 4-9 EADIDACT!VE EFFLUENT MDefITORING SYSTEMS Applicability Applicability Applies to the monitoring systems or programs Applies to the required surveillance of the which perform a surveillance, protective or monitoring systems or programs which perform a controlling function on the release of radioactive surveillance, protective or controlling function cffluents f rom the plant and their identification on the release of radioactive affluents from the in the environment.
P ant and their identification in the environment.
l Objective Objective 4
To assure the operability of the radioactive To specify the type and f requency of surveillance affluent monitoring systems and environmental to be applied to the ridioactive effluent
- programs, monitoring system and environmental programs.
Specifications Specifications A.
Liquid Ef fluent Instrumentation A..
Liquid Effluent Instrumentation 1.
During periods of release through the 1.
Each radioactive liquid effluent monitored pathway, the radioactive monitoring instrumentation channel shall liquid ef fluent monitoring be tested and calibrated as indicated in instrumentation channel shall be Table 4.9.1.
1 operable in accordance with Table 3.9.1 with their alara setpoints set to ensure t hat the limits of Specification 3.8.A.1 I
are not exceeded.
4 Amendeant No.
00 161
8
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VYNPS 3.9 LIMITING CONDITIONS FOR OPERATION 4.9 SURVEILI.ANCE REQUIREMENTS 3.
Caseous Effluent Instrumentation B.
Caseous Effluent Inctrumentation 1.
The gaseous process and effluent 1.
Each gaseous process or effluent monitoring instrumentation channels monitoring instrumentation channel shall shall be operable in accordance with be tested and calibrated as indicated in Table 3.9.2 with their alare/ trip Table 4.9.2.
setpoints set to ensure that the limits of Specifications 3.8.E.1.a. 3.8.J.1, and 3.8.K.1 are not exceeded.
C.
Radiological Environmental Monitoring Program C.
Radiological Environmental Monitoring Program 1.
The radiological envirorssental 1.
The" radiological environmental monitoring program shall be conducted as monitoring samples shall be collected specified in Table 3.9.3.
pursuant to Table 3.9.3 from the locations given in the ODCM and shall be analyzed pursuant to the requirements of Table 3.9.3 and the detection capabilities required by Table 4.9.3.
i G
e Annadment No.
83 161a 1
6 h
VYMPS
-3.9 LIMITING CONDITIONS FOR OPERATION 4.9 SURVEILLANCE REQUIREMENTS n
D.
Land Use Census D.
Land Use Census 1.
A land use census shall be conducted to 1.
The land use census shall be conducted identify the location of the nearest at least once per year between the-dates -
milk animal and the nearest residence in of June 1 and October 1 by either a each of the 16 mete'orological sectors door-to-door survey, serial survey, or within a distance of five miles. The by consulting local agricultural
, i, survey shall also identify the nearest authorities. The results of the land milk animal (within 3 miles of the use census shall be included in the plant) to the point of predicted highest annual Radiological Environmental annual average D/Q value in each of the Surveillance Report pursuant to three major meteorological sectors due Specification 6.7.C.3.
to elevated releases from the plant stack.
2.
With a land use census identifying one or more locations which yield a calculated dose or dose commitment (via the same exposure pathway) at least 20 percent greater than at a location f rom which samples are currently being f
obtained in accordance with Specification 3.9.C.1, add the new location (s) to the radiological A :;endrent No.
83 162 6
VVWra
~
3.9 LIMITING CONDITIONS FOR OPERATION 4.9 SURVEILLANCE REQUIRENENTS environmental monitoring program within 30 days if permission f rom the owner to collect samples can be obtained, and sufficient sample volume is available.
The sempling location (s), excluding the control station location, having the lowest calculated dose or dose commitment (via tha same exposure pathway) may be deleted from this monitoring program af ter October 31 of the year in which this land use census was conducted.
E.
Intercomparison Program E.
Intercomparison Program 1.
Analyses shall be performed on 1.
A summary of the results of analyses referenced radioactive materials performed as part of the above required supplied as part of an Intercomparison Intercomparison Program shall be Program which has been approved by HRC.
included in the Annual Radiological Environmental Surveillance Report. The identification of the NRC approved Intercomparison Program which is being Participated in shall be stated in the ODCH.
AmsndE3nt No.
83 163 i
'l VYNPS TABLE 3.9.1 RADIOACTIVE LIQUID EFFLUENT HONITORING INSTRUNENTATION Minimum Channels Operable Notes 1.
Gross Radioactivity Honitors not Providing Automatic Termination of Release a.
Liquid Radwaste Discharge 1*
1,4,5 Monitor b.
Service Water Discharge 1
2,4,5 Monitor 2.
Flow Rate Measurement Devices a.
Liquid Radwaste Discharge 1*
3,4 Flow Rate'Honitor n
- During releases via this ' pathway.
Amendrent Nog 83 164
l VYNPS TABLE 3.9.1 1
(continued)
TABLE NOTATION With the number of channels operable less than required by the minimum channels operable requirement.
NOTE 1 effluent releases may continue provided that prior to initiating a releases i
I s.
At least two independent samples are analyzed in accordance with Specification 4.8.A.1, and b.
At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving.
Otherwise, suspend release of radioactive effluents via this pathway.
NOTE 2 - With the number of channels operable less than required by the minimum channels operable requirement, i :
effluent releases via this pathway may continue provided that, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 microcurie /ml.
NOTE 3 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves may be used to estimate flow, a
With the number of channels operable less than required by the minimum channels operable requirement, exert NOTE 4 reasonable ef forts to return the instrument (s) to operable status prior to the next release.
NOTE 5 - The alarm setpoints of these channels shall be determined and adjusted in accordance with the methodology l
and parameters in the Of f-Site Dose Calculation Manual (ODCM). With a radioactive liquid effluent i monitoring instrumentation channel alarm setpoint less conservative than a value which will ensure that the l
limits of e.8.A.1 are met during periods of release, immediately take action to suspend the release of radioactive liquid ef fluents monitored by the.affected channel or declare the channel inoperable; or change
(
the setpoint so it is acceptably conservative.
Amendment tio.
80 165 l
l l.
i VYNPS TABLE 3.9.2 CASE 00S EFFLUENT HONITORING INSTRUMENTATION Inst rument Minimum Channels Operable Notes i
1.
Stcan Jet Air Ejector (SJAE) 1 7,8,9 I
c.
Noble Gas Activity Monitor 2.
Augmented Of f-Cas System o.
Noble Gas Activity Honitor Between the Charcoal Bed 1
2,5,6,7 System and the Plant Stack (Providing Alara and Automatic Termination of Release) b.
Flow Rate Monitor 1
1.5,6 c.
Hydrogen Monitor 1
3,5,6 3.
Pirnt Stack c.
Noble Gas Activity Honitor 1
2,5,7 b.
Iodine Sampler Cartridge 1
4,5 c.
Particulate Sampler Filter 1
4,5 d.
Sampler Flow Integrator 1
1,5 o.
Stack Flow Rate Monitor 1
1,5 Atsndnant No.
83 166
}
I e
e.
VYNPS TABLE 3.9.2 (continued)
TABLE NOTATION NOTE 1 With the number of channels operable lets than required by the minimum channels operable. requirement, ef fluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
NOTE 2 With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue for a period of up to 7 days provided that at least one of the stack monitoring systems is operable and off gas system temperature and pressure are measured continuously.
With the number of channels operable less than required by the minimum channels operable requirement, i
NOTE 3 operation of the A0G System may continue provided gas samples are collected at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or a-orderly transfer of the of f gas ef fluents f rom the operating recombiner to the st'andby recombiner shall la made.
, NOTE 4 - With the number of channels operable less than required by the minimum channels operable requirement, effluent release.a via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment.
)
[ NOTE 3 With the number of channels operable lesq than required by the minimum channels operable requirement, exert reasonable efforts to return the instrument (s) to operable status within 30 days.
NOTE 6 During releases via this pathway.
)
NOTE 7 - The alare/ trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the Off-Site Dose Calculation Haaual (ODCM). With a gaseous process or ef fluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of 3.8.E.1.a and 3.8.K.1 are met, immediately take actions to suspend the release of radioactive gaseous effluents ronitored by the affected channel, or declare the channel inoperable, or chande the setpoint so it is acceptably conservative.
Amendmant No.
00 167 I
. s VYNPS TABLE 3.9.2 (continued)
TABLE NOTATION Nsto 8 - Hinimum channels operable required only during operation of the Stean Jet Air. Ejector.
Nsto 9 - With the number of channels operable less than required by the minimum channels operable requirement, gases f ros the SJAE may be released to the environment for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided:
1.
The A0G system is not bypassed; and 2.
The AOG eystem noble gas activity monitor is operable.
9 A: ;adasnt No.
83 168
a.
A dl>
Os VYNFS S
TABLE 3.9.3 RAD 101.0CICAL ENVIRONMENTAL NONITORIMC PROGRAM S.
(
l Esposure Fathway O
a r.d /o r S ample Number of Sample Locatione, Samplig and Collection Frequency Type and Frequency of Analyste 1.
AIRSORNE S
a.
BaJtotodine and Samples from S locationes Continuous operation of esepter Radiotodine canister Analyse particulates with eseple collection seateonthly each eseple for 1-131.
9 3 eseples f rom close to the 3 or more f requently as required by site boundary locatione in duet loading or plant effluent particulate esepler Crose beta h
dif ferent sectore, of the highest releasee.
radioactivity analyste on each sample 3
calculated annual average ground following filter change.C i
level Diq.
Composite (by location) for gamma d
tootopte at least once per quarter.
e 1 mm.sple from the vicipity of a community havius the highest calculated annual average ground 4
level D/Q.
I esople from a control location.
3 se for example 15-30 km distant and in the least prevalent wind direction.
g i
b e
i Amendment No.
dS 4
169 l
v i
i e
1 l
t 4
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t 4
4 4
4 6
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9 9
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- ~23 1
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VYliFS TABLE 3.9.3 (continued)
RADIOLOGICA1. ENVIRONMENTAL NOMITORING pBOGRAM Esposure Fathway and/or Sample alumber of Sample 14 cations, Samplina and Co11ection Freemency Type and Freeuemer of Analyste 3.
WATEasoaWE d of e.
Surface
- 1 eseple upstress.
Heathly grab sample.
Gaana lectopic analyele each sample. Trittua analysis of 1 esaple downstress.
Composite sample coll ted over composite esople at least once per a period of one month quarter.
b.
Cround 1 eseple from withia 8 km Quarterly.
Gaana tootopied med tritius diatsace.
analyses of each sample.
1 sample f rom a control location.
Quarterly.
c.
Sediment from 1 esople f roe downstress area Seateanually.
Cama footopic caatysted of Shore 11ae with esteting or potential each semple, recreational value.
1 eenple from north store drata As specified la outfall.
the 0001.
a a
Amendment No.
Ob 171 i
e e
I l
t e
e
i
=
s.
an 4
A V1NpS A
TABLE 3.9.3 (contImued)
RADIOLOGICAI. ENVIRONMENTAL MONITORINC PROCRAN e
Esposure Fathway anJ/or Sample aanber of Sample Locat ione, Samplina and Collection pr euency Type and Frequency of Analrata g
4.
INGESTION e
s.
Milk Samples free allking animals in Seafsonthly if milking antaals Canes footopicd and 1-131 analysis 3 locations within 5 km distance are identified on pastures at of each sample.
having the highest dose potential.
least once per month at other g
if there are less than 3 primary times.
locations avattable then 1 or more secondary sample from milking O
animals in each of 3 areas between 5 to a km distance where doses are calculated to be N
greater than 1 mres per year.
1 sample f rom attking animals in O
a control location.
b.
Fish 1 eaaple of two recreationally Sestannually.
Comma isotopic analysted on edible O
taportant species in vietatty of Plant discharge area.
portions.
1 sample (preferably of same 4
spectee) in areas not influenced 3
j by plant discharge.
O c.
Vegetation 1 grasa sample at each air Quarterly when available.
Comma tootopic analysted of each sampling station.
sample.
1 silage sample at each allk At time of harvest.
samp!!ng station (as available).
Camma, isotopic analyels0 of eeth sample.
AmenJoeot No.
bb 172 w
U W
l l
1
l l
l VYNPS l
TABLg 3.9.3 (continued)
TABLg NOTATION i o Sp:cific parameters of distance and direction sector f rom the centerline of the reactor and additional descriptions where pertinent, shall be provided for each and every sample location in Table 3.9.3 in a table and figure (s) in the OUCH. Deviations are permitted f rom the required sampling schedule if specimens are unobtainable due to hazardous j
candit io ns, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If i
cpacimens are unobtainable due to sampling equipment malfunction, every reasonable effort shall be made to complete i
corrective action prior to the end of the next sampling period. All deviations f rom the sampling schedule shall be documented in the annual Radiological Environmental Surveillance Report pursuant to Specification 6.7.C.3.
It is rceegnized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time.
In these instances, suitable alternative media and locations may be ch2sen for the particular pathway in question and appropriate substitutions made within 30 da s in the radiological
/
savironmental monitoring program.
In lieu of a Licensee gvent Report and pursuant to Specification 6.7.C.1, idsntify the cause of the unavailability of samples for that pathway and identify the new location (s) for obtaining
' replacement samples in the next Semiannual Radioactive gffluent Release Report and also include in the report a ravised figure (s) and table for the ODCH reflecting the new location (s).
b On2 or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously coy, be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a Tharmoluminescent Dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered co two or more dosimeters. Film badges shall nog be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The f requency of analyais or readour for TLD systems will depend upon the l
chore:teristics of the specific system used and shauld he t-lected r.c obtain optimum dose information with minimal fading.
c Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more af ter sampling t o allow ict radon and thoron daughter decay.
If gross beta activity in air particulate sasples is greater than ten l
tices the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual pamples.
t b
Arendment No.
83 172a I
I e
e.
s VYNPS TABLE 3.9.3 (continued)
TABLE NOTATION d Games isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the effluents from the facility.
O The " upstream sample" shall tc taken at a distance bryond significant influence of the discharge. The " downstream" ecaplo shall be taken in an area beyond but near the mixing zone.
~
f r,apecite as:aple al tquots shall be collected at time intervals that are very short (e.g., hourly) reintive to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.
g E ch meteorological sector shall have an established " inner" and an " outer" monitoring location based on ease of rccuvery (i.e., response time) and year-round accessibility.
h Samplo collection will be performed weekly whenever the main plant stack effluent release rate of I-131, as dster=ined by the sampling and analysis program of Table 4.8.2, is equal to or greater than 1 x 10-1 uC1/sec.
Sample collection will revert back to semimonthly no sooner than at least two weeks af ter the plant stack ef fluent roleste rute of I-131 falls and remains below I x 10-1 uC1/sec.
I l
l i
lAmendssnt No.
83 172b l
1 VYNPS TABLE 3.9.4 I
REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES (a)
Reporting Levels i
Water Airborne Particulate Fish Milk Vegetation l
Analysi s (pC1/1) or Cases (pC1/m3)
(pC1/Kg, wet)
(pC1/11 (pC1/K, wet) 3 H-3 2 x 10 (b) 4 Ma-54 1 x 10 3 x 10 Fe-59 4 x 10 1 x 10 Co-58 1 x 10 3 x 10 4
Co-60 3 x 10 1 x 10 Zn-65 3 x 10 2 x 10 Zr-Nb-95 4 x 10 2
1-131 0.9 3
1 x 10 3
3 Co-134 30 10 1 x 10 60 1 x 10 3
3 Ca-137 50 20 2 x 10 70 2 x 10 2
2 Ba-La-140 2 x 10 3 x 10 i
(c) Reporting levels may be averaged over a calendar quarter. When more than one of the radionuclides in Table 3.9.4 are detected in the sampling medium, the unique reporting requirements are not exercised if the following condition holds:
concentration (1) concentrat ion ( 2) j reporting level (1) + reporting level (2) +...
<1.0.
When radionuclides other than those in Table 3.9.4 are detected and are the result of plant affluents, tha l
potential annual dose to a member of the public must be less than or equal to the calendar year 11mits of Specifications 3.8.B, 3.8.E and 3.8.F.
l (b) Reporting level for drinking water pathways. For nondrinking water pathways, a value of 3 x 104 pC1/1 may be used.
Arendment No.
83 172c
VYNPS TABLE 4.9.1 RADIOACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Instrument Source Instrument Functional Instrument Check Check Calibration.
Test 1.
Gross Radioactivity Monitors not Providing Automatic Termination of Release c.
Liquid Radwaste Discharge Monitor (3)
Once each Prior to Once each Once each day
- each release, 18 months quarter (2) but no more (1) than once each month b.
Service Water Discharge Monitor (3)
One each once each once each once each day month 18 months quarter (2)
(1)
I 3
Flow Rate Measurement Devices t
a.
Liquid Radwaste Discharge Flow Rate Monitor Once each Not Not Once each j
day
- Applicable Applicable quarter
- i
- During releases via this pathway.
l l
l l
Amendrent No.
83 172d
5 VYNFS i
TABLE 4.9.1, i
(continued)
TABLg NOTATION (1)
The Instrument Calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Bureau of Standards) liquid radioactive source positioned in a reproducible geometry with respect to the sensor. These standards shall permit calibrating the system over its normal operating range of energy and rate.
(2)
The Instrument Functional Test shall also demonstrate the Control Room alarm annunciation occurs if any of the following conditions exists:
(a)
Instrument indicate measured levels above the alarm setpoint.
~
(b) Circuit failure.
]
(c) Instrument indicates a downscale failure.
(d) Instrument controle not set in operate mode.
~'
(3)
The alarm setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the Of f-Site Dose Calculation Manual (ODCN).
i 1
I l
i Amendeent No.
00 172e i
t
VYNPS TABLE 4.9.2 CASE 00S EFFLUENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMElffS I
Instrument Instrument Source Inst rument Functional Instrument Check Check Calibration Test 1.
Steam Jet Air Ejector (SJAE) c.
Noble Gas Activity Honitor
- Once each Once each Once each Once each-day **
month 18 months quarter (2)
(3) 2.
Augmented Off gas System o.
Noble Gas Activity Monitor Once each Once each Once each once each day
- month 18 months quarter (1)
(?)
b.
Flow Rate Monitor Once each Not Once each Not day
- Applicable 18 months Applicable a
c.
ti; Jrogen 11onitor Once each Not Once each once each day
- Applicable quarter (4) month 3.
Plant Stack c.
Noble Gas Activity Monitor Once each Once each Once each Once each day month 18 months quarter (2)
(3) b.
Sampler Flow Integrator Once each Not Once each Not week Applicable 18 months Applicable c.
System Flow Rate Monitor Once each Not i
'Not Not day Applicable Applicable 4pplicable During releases via this pathway.
During operation of main condenser SJAE.
Amendernt No.
83 172f t
6-VYNPS
^
i TABLE 4.9.2 (continued)
TABLE NOTATION (1)
The Instrueent Functional Test shall also demonstrate that automatic isolation of this pathway and the Control Room alara annunciation occurs if any of the following conditions exists:
(c) Instrument indicate acasured levels above the alara setpoint.
a (b) Circuit failure.
(c) Instrument indicates a downscale failure.
(d) Instrument controls not set in operate mode.
(2)
The Instrument Functional Test shall also demonstrate that, Control Room alarm annunciation occurs when any of the following conditions exist:
I j
(c)
Instrument indicates measured levels above the alare setpoint.
(b) Circuit failure.
(c) Inst rument indicates a downscale failure.,
(d) Instrument controls are not set in operate mode.
I(3) 1 j
The Instrument Calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Bureau of Standards) radioactive source positioned in 'a reproducible geometry with roepect to the sensor. These standards should permit calibrating the system over its normal operating range of rete capabilities.
\\
l(4)
The Instrument Calibration shall include the use of standard gas samples (high range and low range) containing cuitable concentrations, hydrogen balance nitrogen, for the detection range of interest per Specification 3.8.J.1.
!AmendrentNo.
83 172g t
i f
e
VYNPS TABLE 4.9.3 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (a)(c)(f)
Analysis (d)
Water Airborne Particulate Fish Hilk Vegetation Sediment (pC1/1) or Cas (pC1/m3)
(pC1/Kg, wet)
(pC1/1)
(pC1/Kg, vet)
(pC1/Kg dry)
Grces beta 4
0.01 H-3 3000 Hn-54 15 130 Fo-59 30 260 co-58,60 15 130 Zn-65 30 260 Zr-Hb-95 15(b) 1-131 0.07 1
60
\\
Ca-134 15 0.05 130 15 60 150 Co-137 18 0.06 150 18 80 180 83-Lo-140 15(b)(e) 15(b)(e) l Amendesnt No.
00 172h f
~
VYNFS
~
TABLE 4.9.3 (continued)
I TABLE NOTATION i
L See Footnote (a) of Table 4.8.1.
(a)
(b)
Parent only.
If the sensured concentration minua the 5 sigma counting statistics is found to exceed the specified LLD, (c)
' the sample does not have to be analyzed to meet the specified LLD.
(d)
This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, t'ogether with those of the listed nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Surveillance Report pursuant to Specification 6.7.C.3.
The Ba-140 LLD and. concentration can be determined by the analysis of its short-lived daughter product (o)
L -140 subsequent to an 8 day period following collection. The calculation shall be predicted on the Mrmal ingrowth equations for a parent-daughter situation and the assumption that any unsupported La-140 in the sample would have decayed to an insignificant amount (at least 3.6 percent of its original value). The ingrowth equations will assume that the supported La-140 activity at the time of collection is sero.
Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD, but (f) as "not detected". For purposes of averaging, the LLD will be assumed to be zero.
t i
Amendment No.
83 1721 1
i; I
i-VYNPS
'l:
GASES:
3.9 RADIDACTIVE EFFLUENT NONITORING SYSTEMS A.
Liquid Ef fluent Instrumentation The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid ef fluents during actual or potential releases of liquid effluents. The alarm setpoints for these instruments are to ensure that the alarm will occur prior to exceeding the limits of 10CFR Part 20.
~
Automatic isolation function is not provided on the liquid radwaste discharge line due to the inf requent
, nature of batch, discrete volume, liquid discharges (on the order of once per year or less), and the administrative controls provided to ensure that conservative discharge flow rates / dilution flows are set such that the probability of exceeding the 10CFR Part 20 concentration limits are low, and the potential of f site dose consequences are also low.
B.
Gaseous Effluent Instrumentation i
The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous af fluents during actual or potential releases of gaseous effluents. The alare/ trip setpoints for these instruments are provided to ensure that the alare/ trip will occur prior to exceeding the lietts of 10CFR Part 20.-
This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas hoidup system.
1 Amendment No.
00 172j f
~!
WTWP8 1
3.9 (Continued) f 1
C.
Radiological Rnvironmental Monitoring Program 5
i The radiological monitoring program required by this specification provides measurements of radiation and l
of radioactive materials it-those exposure pathways and for those radionuclides which lead to the highest j
potential radiation exposures of member (s) of the public resulting f rom the station operation. This 1
monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the affluent measurements and modeling of the environmental exposure pathways.
J Ten years of plant operation, including tF: years prio-to the implementation of the Augmented Off-Gas System, have amply demonstrated via routine ef fluent and environmental reports that plant ef fluent seasurements and modeling of environmental pathways are adequately conservative. In all cases, 1
environmental sample results have been two to three orders of magnitude less than expected by the model
)
employed, thereby representing small percentages of the ALARA and environmental reporting levels. This radiological environmental monitoring program has therefore been significantly modified as provided for by 3
Regulatory Guides 4.3 (C.2.a) and 4.1 (C.2.b), Revision 1, April 1975. Specifically, the air particulate and radioiodine air sampling periods have been increased to semimonthly, based on plant effluent and environmental a1 sampling data for the previous ten years of operation. An 1 131 release rate trigger value of 1 x 10~{ uC1/sec f rom the plant stack will require that air sample collection be increaged to veekly. The 1 x 10-1 uCL/see I-131 value corresponds to-the LLD air concentration of 0.07 pct /m at the aguimum predicted air monitoring station, which exhibits a maximum quarterly I/Q value of 2 x 10-7 sec/m. A factor of 3.5 below the Lug value has also been included in the stack release rate value to account for meteorological fluctuations in I/Q. Due to the large local population of cows and the ready availability of milk samples, food product sampling has been eliminated f rom the program in lieu of milk j
sampling. Since milking cows in the area spend very little time on pasture, silage and grass sampling have j
been instituted as an indicator of radionuclide deposition.
l l
The detection capabilities required by Table 4.9.3 are considered optimum for routine environmental measurements in industrial laboratories.
It should be recognized that the LLD is defined as,a before-the-fact limit representing the capability of a measurement system and not as an af ter-the-f act timit for a particular measurement. This does not preclude the calculation of an af ter-the-fact LLD for a particular measurement based upon the actual parameters for the sample in question.
Amendment No.
83 172k o
D h
VYNPS t
i 3.9 (Continued)
D.
Land Use Census This specification is provided to ensure that changes in the use of areas at and beyond the site boundaries are identified and that modifications to the monitoring program are made if reqaired by the results of this I
census. This census satisfies the requirements of Section IV.B.3 of Appendix 1 to 10 CFR Part 50.
The i
requirement of a garden census has been eliminated along with the food product monitoring requirement due to the substantial and widespread occurrence of dairy farming in the surrounding area which dominates the food uptake pathway.
The addition of new sampling locations to Specification 3.9.C, based on the land use census, is limited to those locations which yield a calculated dose or dose commitment greater than 20 percent of the calculated dose or dose commitment at any location currently being sampled. This eliminates the unnecessary changing of the environmental radiation monitoring program for new locations which, within the accuracy of the calculation, contributes essentially the same to the deze or dose commitment as the location already sampled. The substitution of a new sampling point for one already sampled when the calculated dif ference in dose is less than 20 percent, would not be expected to result in a significant increase in the ability-to detect plant effluent related nuclides.
g.
Intercomparison Program
[
J The requirement for participation in an intercomparison program is provided to ensure that independent checks on the precision and securacy of the measurc aents of radioactive material in environmental sample matrices are performed as part of a qual}ty assuraace program for environmental monitoring in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix 1 t'a 10 l
CFR Part 50.
1 4
i,
)
Amendment No.
00 1721 i
l l
VYNPS P
3.11 LIMITING 00NDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS C.
Minimum Critical Power Ratio (dCPR) 1.
During steady state power operation, the MCPR Operating Limit shall be equal or greater than the values shown on Table 3.11-2.
For core flows other than rated MCPR, the Operating NCPR Limit shall be the above value multiplied by Kg g
where Kf is given-by Figure 3.11-2.
If at any I
. time during steady state opera *,.on it is determined by normal surveillance that the Ifmiting value for.
HCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the steady-state MCPR is i
not returned to within the prescribed limits within two (2) hours, the reactor power shall be brought to shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance 5,
and corresponding action shall continue until reactor operation is within the prescribed limits.
6 l
l l
Amendment No. g g3 T80c I
e B
5-t VYNFS Dmas 3.11C Minimus Critical Power Ratio (MCPR)
Operating Limit MCPR 1.
The MCPR Operating Limit is a cycle-dependent parameter which can be determined for a number of different combinations of operating modes, initial conditions, and cycle exposures in order to provide reasonable assurance against exceeding the Fuel Ciadding Integrity Safety Limit (FCISL) for potential abnormal occurrences. The MCPR operating limits are presented in Appendix A of the l
current cycle'a Core Performance Analysis report.
)
a 4
Madtent No. 41, J0' 83 180h 1
-o
~-
VYNFS j
5.0 DESIGN FEATURES 5.1 Site The station is located on the property on the west bank of the Connecticut Riv6r in the Town of Vernon, I
Vermont, which the Vermont Yankee Nuclear Fower Corporation either owns or to which it has perpetual rights and easements. The site plan showing the exclusion area boundary, boundary for gaseous ef fluents and boundary for liquid effluents is on Figire 2.2-5 in the FSAR. The minimum distance to the boundary of the exclusion area as defined in 10CFR100.3 is 910 feet.
No part of the site shall be sold or leased and no structure shall be loqated on the site except structures owned by the Vermont Yankee Nuclear Fower Corporation or related utility companies and used in conjunction with normal utility operr.tions.
5.2 Reactor A.
The core shall consist of not more than 368 fuel assemblies.
j 1
B.
The reactor core shall contain 89 cruciform-shaped control rods. The control material shall be boron carbide powder (8 C).
4 5.3 Reactor Vessel The reactor vessel shall be as described in Table 4.2-3 of L'ae FSAR. The applicable design codes shall be 1
as described in subsection 4.2 of the,FSAR.
5.4 Containment A.
The principal design parameters and applicable design codes for the primary containment shall be as given in Table 5.2.1 of the FSAR.
B.
The secondary containment shall be as described in subsection 5.3 of the FSAR and the applicable codes shall be as described in Section 12.0 of the FSAR.
C.
Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in subsection 5.2 of the FSAR.
Amendment No.
03 188 i
VYNPS i
d.
Power Plant Design l
e.
Reactor Engineering f.
Radiation Safety 3
Solety Aaalysis h.
Instrumentation and Control i.
Metallurgy 3.
Meeting Frequency: Semi-annually and as required on cati of the Chairman.
4.
Quorus: Chairman or Vice Chairman plus four members or designated alternates.
5.
Responsibilities:
Review proposed changes to the operating license including Technical Specifications.
a.
Review minutes of meetings of the Plant Operation Review Committee to determine if matters b.
considered by that committee involve unreviewed or unresolved safety questions.
Review the safety evaluations for changes to equipment or systems completed under the c.
provisions of Section 50.59 10 CFR to verify that such actions did not constitute an unreviewed safety qu6stion.
l d.
Periodic audits of implementing procedures, shall be performed under cognizance of the Committee. Included in these audits, but not limited to, are the following specific activities:
P ant operations; l
1.
11.
facility fire pro'tection program; a
197 j
Amendment No. 66, Jdf, 83 1
?
u
VYNPS 111. the radiological environmental monitoring program and the results thereof at least once per 12 months; iv.
the off-Site Dose Calculation Manual and implementing procedures at least once per 24 '
i months; v.
the Process Control Program and implementing procedures for processing and packaging l
of radioactive waste at least once per 24 months; v1.
the performance of activities. required by the Quality Assurance Program to meet the.
provisions of Regulatory Guide 1.21. Revision 1, June 1974, and Regulatory Guide 4.1, Revision 1 April 1975, at least once per 12 months.
e.
Investigate all reported instances of violations of Technical Specifications, reporting i
findings an'd recommendations to prevent recurrence to the Manager of Operations.
l
{
f.
Perform special reviews and investigations and render reports thereon as requested by the i
Manager of Operations.
I 3
Review proposed tests and experiments and results thereof when applicable.
h.
Review abnormal performance of plant equipment and anomalies.
i t
i.
Review unusual occurrences and incidents which are reportable under the provisions of 10 CFR Part 20 and 10 CFY Part 50.
i j.
J.
Review of occurrences if safety limits are exceeded.
l 6.
Authority l
a.
Review proposed changes to the operating license including Technical Specifications 'and revised bases for submittal to the NRC.
b.
Review proposed changes or modifications to plant systems or equipment, provided that such chanaes or modifications do not involve unreviewed safety questions.
l Arenduent No. 66, Jd, 83 198 I
l l
VYNPS c.
Recommend to the Manager of Operations appropriate action to prevent recurrence of any violations of Technical Specifications.
f d.
Evaluate actions taken by the Plant Operation Review Committee.
7.
Records Minutes of all meetings of this committee shall be recorded. Copies of the minutes shall be forwarded to the Manager of Operations, the Vice President - Operations, the Plant Manager and l
any others that the Chairman may designate.
1 J
i i
i
...e i
Amndeant No.
83
- 193,
?
r o
VYNPS i
6.5 Plant Operating Procedures A
Detailed written procedures, involving both nuclear and non nuclear safety, including applicable check-off lists and instructions, covering areas listed below shall be prepared and approved.
All procedures shall be adhered to.
1.
Wormal startup, operation and shutdown of systems and components of the facility.
2.
Refueling operations.
3.
Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, suspected Primary System leaks and abnormal reactivity changes.
4.
guergency conditions involving potential or actual release of radioactivity.
5.
Preventive and corrective maintenance operations which could have an effect on the safety of the reactor.
4 6.
Surveillance and testing requirements.
4 7.
Fire protection program implementation including minimum fire brigade requirements and training.
l The training program shall eget or exceed the requirements of Section 27 of the NFPA Code 1976.
Training sessions will be scheduled as plant operations permit but will be completed in specified 2
subjects annually.
Initial fire brigade training shall be completed by March 13, 1978.
8.
Process Control Program in plant implementation.
9.
Off-Site Dose Calculation Manual in plant implementation.
I l
I P
Amendment No. 36, 42, kT, 83 200 i
4 1
i i
f
1 VYNPS
~
i B.
Radiation control standards and procedures shall be prepared, approved and maintained and made i
available to all station personnel. These procedures shall show permissible radiation exposure, and shall be consistent with the requirements of 10 CFR Part 20.
This ra,diation prokection program shall i
be organized to meet the requirements of 10 CFR Part 20.
1.
Paragraph 20.203, " Caution signs, labels, signals and controls". In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2), each high radiation area in which the j
intensity of radiation is 1000 mres/hr or less shall be barricaded and conspicuously posted as a high radiation area and ent rance thereto shall be controlled by requiring issuance of a Radiation Work Permit.* Any individual or group of individuals permitted to enter such areas shall be provided with one or more of the followings a.
A radiation monitoring device which continuously indicates the radiation dose rate in the area.
b.
A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. gntry into such areas with this monitoring device may be made af ter the dose rate levels in the area have been established and personnel have been made knowledgeable of them.
i c.
A Health Physics qualified individual (i.e., qualified in radiation protection procedures)
I with a radiation dosa rate monitoring device who is responsible for providing positive control over the activities within the area and who will perform periodic radiation' surveillance at the f requency specified in the RWP. The surveillance f requency will be established by the Plan ( Health Physicist.
The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mees /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative l
control of the Shif t Supervisor on duty and/or the Plant Health Physicist.
! 8Haalth Physica personnel shall be exempt f rom the RWP issuance requirement during the performance of their assigned i
rcdiction protection duties, providing they are following plant radiation protection procedures for entry into high i
radiation areas.
l Anandment No. 36, kf, d3 201 4
i.
j i
i 9
VY til'S l C.
Procedures prepared fo r A and B above shalI be reviewed and approved by the Plant Hanager, or his designee, and the flanager of Operations.
Tamporary changes to procedures described in Specification 6.5.A above wlilch do n.ot change the intent of the D.
Such original procedure, may be made with the concurrence of two Individuals holding senior operator licenses.
changes shall be documented and subsequently reviewed by the PORC and approved by the Plant Manager or his designee.
Tamporary changes to procedures described in Specification 6.5.B may be made with the concurrence of an individual E.
holding a senior operator license and the healtli physicist on duty.
F.
Licensed radioactive sealed sources shall be leak tested f or contamination. Tests f or leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the Commission or en agreement state as f ollows:
1.
Each IIcensed sealed source, except startup sources previously subj~ected to core flux, containing radloactive materials, other than Ilydrcgen 3, with half-life greater than thirty days and in any form, other than gas, shall be tested f or leakage and/or contamination at intervals not to exceed six isontles.
2.
'Ihe periodic leak tes t required does not apply to sealed sources that are stored and are not being used. The sources exempted f rom t his t est shall be tested f or leakage prior to any use or t ransf er to another user unless they have been leak tested within six months prior to the date of use or transfer.
In the absence of a certificate f rom a transferrer indicating that a leak test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.
3.
Each scaled startup sontce shall be tested within 31 days prior to being subjected to core flux and following repal r or maintenance to t he source.
'the leakage t est; shall be capable of detecting the presence of 0.005, microcurie of radioactive material on the test sample.
If the test reveals the presence of 0.005 microcurie or store of removable contamination, it shall levnediately be withdrawn f rom us, decontaminated, and repaired, or be disposed of in accordance with Commission regulations.
flotwit hstanding the periodic leak test s required by this Technical Specification, any licensed sealed source is exempt f rom s.uch leak test when the source contains 100 microcuries or less of beta and/or gamma emitting material or 5 cicrocuries or less of alpha esiitting material.
A speclat report shall be prepared an l submitted to the Commission within 9b days i f source leakage tests reveal the presence of >0.005 microcurles of removable contaminution.
Amendment flo. D8' 83 20 t, t
1 TanPL 7.
Records of transient or operational cycling for those plant components that have been designed to J
operate safely for a limited number of transients or operational cycles.
8.
Records of inservice inspections of the reactor coolant system.
9.
Minutes of meetings of the Plant Operation Review Committee and the Nuclear Safety Audit and Review Board.
10.
Records for Environmental Qualification which are covered under the provisions of paragraph 6.9.
I 11.
Records of analysis required by the Radiological Environmental Monitoring Program.
G o
l Amendzent No. 31, Order dated October 24, 1980 43 3 207a i
8 I
VYNPS 6.7. Reportina Requirements In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwise noted.
A.
Routine Reports T
1.
Startup Report A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a dif ferent design or has been manufactured by a j
dif ferent fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall, in general, include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this re po r t.
Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption of commencement of commercisl power operation, or (3) 9 months following initfA1 Criticality, whichever is earliest.
If the startup report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports aball be submitted at least every three months until all three events have been completed.
i Amendaent No, )( d3 208 9
f.
~
VYNPS 2.
Annual Report I.
An annual report covering the previous calendar year shall be submitted prior to March I of each year. The annual report shall include a tabulation on an annual basis of the nunber of station,
\\
utility and other personnel (including contractors) _ receiving exposures greater than 100 mren/yr and their associated man-res exposure according to work and job functions,1/ e.g.,
reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD or film badge measurement. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be i
assigned to specific major work functions.
3.
Monthly operating Report Roetine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Office of Management Information and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the appropriate Regional Office, to arrive no later than the fif teenth of each month following the calendar month covered by the report. These reports shall include a narrative summary of operating experience during the report period which describes the operation of the facility and any major safety-related maintenance.
B.
Reportable Occurrences Reportable occurrences, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurre nce.
In case of corrected or supplemental reports, a licensee event report shall be completed, and reference shall be made to the original report date. Events involving systems or components described in Sections 3/4.8.B, 3/4.8.C 3/4.8.F 3/4.8.C, 3/4.8.H. 3/4.8.I 3/4.8.M, 3/4.9.C, 3/4.9.D, 3/4.9.E, Table 3.9.1-note 5, Table 3.9.2-note 7 and 3/4.13 do not require reporting under the provision of this section. Such events will be reported as required in Section 6.7.C.2 or 6.7.C.3 as indicated below. The reporting provisions of this section are not applicable to Sections 3/4.8.A, 3/4.8.D., 3/4.8.E, 3/4.8 N. 3/4.9. A, and 3/4.9.B.
l If This tabulation supplements the requirements of 20.407 of IDCFR Part 20 l
l Amendment No. A1, 83 209 1
i VYNPS I
I 1.
Prompt Notification With Written Follow-Up I
The types of events listed below shall be reported as expeditiously as possible, but within 24
~
hours by telephone and confirmed by telegraph, mailgras, or facetalle transmission to the Director of the appropriate Regional Office, or his designate, no later than the first working day following the event, with a written follow up report within two weeks. The written follow up report shall include, as a minimum, a completed copy of a licensee event report form.
Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provi,de complete explanation of the circumstances surrounding the event.
a.
Failure of the Reactor Protection System or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the Technical specifications or fatture to complete the required protective function.
Note: Instrument drif t discovered as a result of testing need not be reported under this ites but may be reportable under Items 1.e 1.f or 2.a below, b.
Operation of the unit or affected systems when any parameter or operation subject to a limiting condition is less conservative than the least conservative aspect of the limiting
{
condition for operation established in the Technical Specifications.
4 Note: If specified action is taken when a system is found to be operati,ng between the most conservative and,the least conservative aspects of a limiting condition for operation listed in the Technical Specifications, the limiting condition for operation is not considered to have been violated and need not be reported under this ites, but it may be reportable under Ites 2.b below.
Abnormal degradation discovered in fuel cladding, reactor coolast pressure boundary or c.
Note Leakage of valve packing or gaskets within the limits for identified leakage set forth in Technical Specifications need not be reported under this ites.
Assadsant No.
63 210 k
i
4
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VYNPS i
d.
Reactivity anomalies, involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation, greater than or equal to II k/k; a calculated reactivity balance indicating a shutdown margin less conservative than specified in the Technical Specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, of subcritical, an unplanned reactivity insertion of more than 0.5% ok/k or occurrence of any unplanned criticality.
e.
Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of systes(s) used to cope with accidents analyzed in the SAR.
f.
Personnel error or procedural inadequacy which prevents or caiuld prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.
'i k
Note For items 1.e and 1.f. reduced redundancy that does not result in a loss of system function need not be reported under this section bat may be reportable under Items 2.b and 2.c below.
g.
Conditions arising f rom natural or man-made events that, as a direct result of the event, require plant shutdown, operation of safety systems, or other protective measures required by Technical Specifications.
h.
Errors discovered in the transient or accident analyses or in the methods used for such i
analyses as described 1:3 the safety analysis report or in the bases for the Technical
)
Specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.
l l
1.
Performance of structures, systems or components that requires remedial action or l
corrective seasures to prevent operation in a manner less conservative than assumed in the l
accident analyses in the safety analysis report or Technical Specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report or Technical Specifications that require remedisl action or corrective measures to prevant the aristence of develope.es: of an unsafe condition.
Note: This item is intended to provide for reporting of potentially generic problems.
Amend 2ent No.
d0 211 l
l
i VYNPS l
2.
Thirty Day Urf tten Reports The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional Of fice within 30 days of occurrence of the event. The written report shall include, as a minimum, a completed. copy of a licensee event report form.
Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
a.
Reactor Protection System or engineered safety feature inst rument settings which are found to be less conservative than those established by the Technical Specifications but which do not prevent the fulfillment of the functions 1 requirements of affected systems.
b.
Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.
Note: Routine surveillance testing, instrument calibration or preventative maintenance which require system configurations as described in Items 2.s and 2.b need not be reported except where' test results themselves reveal a degraded mode as described above.
c.
Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in Reactor Protection Systems or Engineered Safety Feature Systems.
d.
Abnarmal degradation of systems other than those specified in Item 2.c above designed to contain radioactive material resulting f rom the fission process.
Note: Sealed saurces or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in i
Technical Specifications need not be reported under this item.
?
Am:nd :nt No.
63 212 l
l
VYNPS I
C.
Unique Reporting RequLrements 5
1.
Semiannual Effluent Release Report a.
Within 60 days af ter January 1 and July 1 of each year, a report shall be submitted covering the radioactive content of effluents released to unrestricted areas during the previous six months of operation.
b.
The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released f rom the unit as outlined in Regulatory Guide 1.21, Revision 1. June 1974, " Measuring Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Ef fluents f rom Light-Water-Cooled Nuclear Power Plants", with data summarised on a quarterly basis following the format of Appendix B thereof. For solid wastes the format for Table 3 in Appendix B of Regulatory Guide 1.21"shall be supplemented with three additional categories: class of solid wastes (as defined by 10CFR Part 61), type of container (e.g., LSA, Type A, Type 5, Large Quantity), and solidification agent or absor be nt, if any.
In addition, the radioactive effluent release report to be submitted 60 days af ter January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour a
listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atpospheric stability
- This same report (or a supplement to it to be e
submitted within 180 days of January 1 each year) shall include an assessment of the radiation doses due to the radioactive 11guld and gaseous ef fluents released f rom the unit during the previous calendar year. (The semiannual effluent release report submitted within 60 days of July 1 each year need not contain any dose estimates f rom the previous 6 months' effluent releases.) The effluent reported submitted after January 1 each year shall also include an assessment of the radiation doses f rom radioactive affluents to
- In lieu of submission with the first half year radioactive ef fluent release report, the licensee has the options of retaining this summary of required meteorological data in a file that shall be provided to the NRC upon request.
Amendment No.
83 213 lI
VYNPS '
N member (s) of the public due to any allowed recreational activities inside the site boundary during the previous calendar year. All assumptions used in making these assessments (e.g.,
j specific activity, exposure time and location) shall be included in these reports. For any batch or discrete gas volume releases, the meteorological conditions concurrent with the time of release of radioactive materials in gaseous ef fluents (as determined by sampling f requency and measurement) shall be used for determining the gaseous pathway doses. For radioactive materials released in continuous effluent streams, quarterly average meteorological conditions concurrent with the quarterly release period shall be used for l
determining the gaseous pathway doses. The assessment of radiation doses shall be l
performed in accordance with the Of f-Site Dose Calculation Manual (ODCN).
With the limits of Specification 3.3.N.1 being exceeded during the calendar year, the radioactive ef fluent release report to be submitted 60 days af ter January 1 of each year shall also include an assessment of radiation doses to the likely most exposed real member (s) of the public f rom reactor releases (including doses f rom primary effluent pathuays and direct radiation) for the previous calendar year to show conformance with l
40CFR190, Environmental Radiation Protection Standards for Nuclear Power Operation.
The radioactive effluent release reports shall include a list and description of unplanned releases f rom the site to site boundary of radioactive materials in gaseous and liquid effluents made during the reporting period.
With the quantity of radioactive material in any outside tank exceeding the limit of Specification 3.8.D.1, describe the events leading to this condition in the next Radioactive Ef fluent Eglease Report.
If inoperable radioactive liquid effluent monitoring instrumentation is not returned to operable status prior to the next release pursuant to Note 4 of Table 3.9.1, explain in the next Radioactive Effluent Release Report the reason (s) for delay in correcting the inoperability.
If inoperable gaseous effluent monitoring instrumentation is not returned to operable
~
status within 30 days pursuant to Note 5 of Table 3.9.2, explain in the next Radioactive Effluent Release Report the reason (s) for delay in correcting the inoperability.
214 Amendnint No.
g3 j
VYMPS With allk samples no longer available f rom one or more of the sample locations required by Table 3.9.3, identif y the cause(s) of the sample (s) no longer being available, identify the new location (s) for obtaining available replacement samples, and include revised ODCM figure (s) and table (s) reflecting the new location (s) in the next Radioactive Effluent Release Report.
With a land use census identifying one or more locations which yield at least a 20 percent greater dose or dose commitment than the values currently being calculated in Specification 4.8.G 1, identify the new location (s) in the next Radioactive Effluent Release Report.
Changes made during the reporting period to the Process Control Program (PCF) and to the f
Off-Site Dose Calculation Manual (ODCH), shall be identified in the next Radioactive I
Ef fluent Release Report.
2.
Special Reports
~
Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report.
Liquid Effluents, Specifications 3.8.B and 3.8.C.
a.
With the calculated dose f rom the release of radioactive materials in 11guld af fluents q
exceeding any of the limits of Specification 3.8.5.1, prepara and submit to the Coenission within 30 days a special report which identifies the cause(s) for exceeding the limit (s) and defines the correqtive actions taken to assure that subsequent releases will be in compliance with the limits of Specification 3.8.B.1.
I With liquid radwasta being discharged without processing through appropriate treatment i
systems and estimated doses in exceas of Specification 3.8.C.1, prepare and submit to the Commission within 30 days a apecial report which includes the following information:
l l
(1) explanation of why liquid radweste was being discharged without treatment.
identification of any inoperable equipment or subsystems, and the rea' sons for the l
inoperability;
.g Amend ment No.
00 215 i
t
e VYMPS (2) action (s) taken to restore the icoperable equipment to operable statuel and (3)
Summary description of action (s) taken to prevent a recurrence.
b.
Gaseous affluents, Specifications 3.8.F. 3.8.G, 3.8.H. and 3.8.I.
With the calculated air dose f rom radioactive noble gases in gaseous ef t. 4ents exceeding any or the limits of Specification 3.8.F.1, prepara and submit to the Commission within 30 l
days a special report which identifies the cause(s) for exceeding the limit (s) and the corrective action (s) taken to assure that subsequent retsases will be in compliance with the limits of Specification 3.8.F.1.
With the calculated dose from the release of Iodine-131, Iodine-133, tritium, and/or radionuclides in particulate form exceeding any of the limits of Specification 3.8.G.1, prepara and submit to the Commission within 30 days a special report which identifies the cause(s) for escoeding the limit (s) and the corrective action (s) taken to assure that subsequent releases will be in compliance with the limits of Specification 3.8.G.I.
With gaseous radwaste being discharged without processing through appropriate treatment systems as defined in Specification 3.8.H.1 for more than seven (7) consecutive days, or in excess of the limits of Specification 3.8.1.1, prepara and submit to the Commission within 30 days a special report which includes the following information:
(1) explanation of why gaseous radweste was being discharged without treatment (Specification 3.8 H.1), or with resultant doses in excess of, Specification 3.8.1.1, Identifiestion of any insparsb*e equip 4en, or subsystess, ar.d the reasons for the inoperability; (2) action (s) taken to restore the inoperable equipment to operable status; and (3) summary description of action (s) taken to prevent a recurrence.
i Amendment No. 83 216 i
f i
i I
l VYMPS l
c.
Total Dose, Specification 3.8.M.
With the calculated dose f rom the release of radioactive materials in liquid or gaseous ef fluents exceeding the limits of Specification 3.8.M. prepara and submit to the Commission within 30 days a special report which defines the corrective action (s) to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification i
3.8.M and includes the schedule for achieving conformance with these lietts. This special report, required by 10CFR Part 20.405c shall include an analysis that estimates the radiation exposure (dose) to a member of the public from station sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of j
radioactive material involved, and the cause of the exposure levels or concentrations. If i
the estimated doses exceed any of the limits of Specification 3.8.M. and if the release condition resulting in violation of 40CFR Part 190 has not already been corrected, the l
special report shall include a request for a variance in accordance with the provisions of j
40CFR Part 190. Submittal of the report is considered a timely request, and a variance is j
granted until staff action on the request is complete.
d.
Radiological Environmental Monitoring Specification 3.9.C.
With the lavel of radioactivity as the result of plant effluents in an environmental l
sampling media at one or more of the locations specified in Table 3.9.3 exceeding the reporting levels of Table 3.9.4 prepare and submit to the Commission within 30 days f rom i
the receipt of the Laboratory Analyses a special report which includes an evaluation of any i
release conditions, envi,ronmental fact'o'rs or other factors which caused the limits of Table j
' 3.9.4 to be exceeded. This report is not required if the measured level of radioactivity I
i was not the result of plant effluents, however, in such an event, the condition shall be reported and described in the annual Radiological Environmental Surveillance Report.
g i
I e.
Land Use Census, Specification 3.9.D.
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With a land use census not being conducted as required by Specification 3.9.D., prepare and 1
j submit to the Commission within 30 days a special report which identifies the reasons why the survey was not conducted, and what steps are being taken to correct the situation.
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)Assadter.tNo.
83 217 l
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3.
Environmental Radiological Monitoring Radiological Environmental Surveillance Esports covering the operation of ths unit during previous calendar year shall be submitted prior to May 1 of each year.
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I The annual Radiological Environmental Surveillance Report shall include summaries.
interpretations, and' an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with operational controla l
(as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.
The annual Radiological Environmental Surveillance Report shall include suasarised and tabulated results of all radiological environmental samples taken during the report period pursuant to the -
table and figures in the ODCH.
In the event that some results are not available for inclusion I
with the report, the report shall be submitted noting and explaining the reasons for the missing i
results. The missing data shall be submitted as soon as possible in a supplementary report.
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With the level of radioactivity in an environmental sampling media at one or more of the i
locations specified in Table 3.9.3 exceeding the reporting levels of Table 3.9.4, the condition I
shall be described in the next annual Radiological Environmental Surveillance Report only if the measured level of radioactivity was not the result of plant effluents. With the radiological environmental monitoring progras not being conducted as specified in Table 3.9.3, a description 5
i of the reasons for not conducting the program as required and the plans for preventing a recurrence shall be included in the next annual Radiological Environmental Surveillance Report.
I The annual Radiological Environmental Surveillance Report shall also include the results of the land use census required by Specification 3.9.D.
A summary description of the radiological l
environmental monitoring program including a map of all sampling locations keyed to a table l
giving distances and directions f rom the reactor shall be in the reports. If new environmental i
sampling locations are identified in accordance with Specification 3.9.D. the new locations shall l
be identified in the next annual Radiological Environmental Surveillance Report.
l The reports shall also include a discussion of all analyses in which the LLD required by Table j
4.9.3 was not achievable.
i Amendment No.
33 218 i
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i VYMPS l
The resulta of licensee participation in the intercomparison program required by Specification 3.9.E shall be included in the reports. With analysee not being performed as I
required by Specification 3.9.E. the corrective actions taken to prevent a recurrence shall be l
report to the Commission in the next annual Radiological Environmental Surveillance Report.
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l 6.8 Fire Protection Inspection l
A.
An independent fire protection and loss prevention inspection and audit shall be performed annually i
utilising either quellfied off site licensee personnel or an outside fire protection fire.
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B.
An inspection and audit by an outside fire consultant shall be performed at intervals no greater than 3 years.
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l 6.9 Environmental Qualification i
A.
By no later than June 30, 1982, all safety-related electrical equipment in the facilityeshall be l
l qualified in accordance with the provisions of Division of Operating Reactors, " Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Esectora" (DOR Guidelines); or NUEEG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related l
Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License DPR-28, dated October 24, 1980.
B.
By no later than December 1,1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all' safety related l
electrical equipment in sufficient, detail to document the degree of compliance with the DDR Guidelines j
or NUEEG-0588. Thereaf ter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise qualified.
l 6.10 Integrity of Systems outside Containment i
l A program to reduce leakage f rom systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels will be implemented. This i
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program shall include the following:
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l Amend:ent No.
33 219 i
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VYNPS A.
Provisions establishing preventive maintenance and periodic visual inspection requirements.
l R.
Systes leakage inspections, to the extent permitted by system design and radiological conditions, for each systes at a f requency not to exceed refueling cycle intervals. The systems subject to this testing ares (1) Residual Heat Removal, (2) Core Spray, (3) Reactor Water Cleanup, (4) EPCI,
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(5) RCIC, and (6) Sampling Systems.
6.11 Iodine Monitorina 1
A program which will ensure the capability to accurately determine the airborne lodine concentration in vital areas
- under accident conditions will be implemented. This program shall include the following:
A.
Training of personnel.
5.
Procedures for monitoring.
i C.
Provisions for maintenance of sampling and analysis equipment.
1 6.12 Process Control Progree (PCP)
A process control program shall contain the sampling, analysis, tests, and determinations by which wet radioactive waste f rom liquid systems is assured to be converted to a form suitable for off site disposal.
A.
Licensee initiated changes to the PCP:
1.
Shall be submitted to the Commission in the sentannual Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
s.
Suf ficiently detsited information to support the rationale for the change without benefit of additional or supplemental information.
- Arsos requiring personnel access for establishing hot shutdown condition.
Amendment No.
$3 3
220
VYIIPS b.
A determination that the change did not reduce the overall conformance of the dewatered spent resins / filter media waste product to existing criteria for solid weste shipments and disposal.
c.
Documentation of the fact that the change h'as been reviewed by PORC and approved by the Manager of Operations (H00).
2.
Shall become of fective upon review by PORC and approval by the Manager of Operations (M00).
6.13 Off-Site Dose Calculation Manual (OOCN)
An Off-Site Dose Calculation Manual shall contain the current methodology and parameters used in the calculation of off site doses due to radioactive gaseous and liquid ef fluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the environmental radiological monitoring program.
A.
Licensee initiated changes to the ODCM:
1.
Shall be submitted to the Commission in the semiennual. Rffluent Release Report for the period in which the change (s) was mada ef fective. This subcittal shall contain a.
Sufficiently detailed information to support the rationale for the change without bene, fit of additional or supplemental information.
Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date, box, together with appropriate analyses or evaluations justifying the c hange( s).
b.
A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations.
c.
Documentation of the fact that the change has been reviewed by PORC and approved by the Manager of Operations (H00).
2.
Shall become effective upon review by PORC and approved by the Manager of Operations (M00).
4 03 Amendment No.
s 221 1
i VYNFS 6.14 Major Channes to Radioactive Liquid, Caseous, and Solid Waste Treatment Systeme*
Licensee-initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):
Shall be reported to the Commission in' the semiannual Radioactive Effluent Release Report for the A.
The discussion of each change shall contain period in which the evaluation was reviewed by the PORC.
A summary of the evaluation that led to the determination that the change could be made in 1.
accordance with 10CFR Part 50.59; Suf ficient detailed information to support the reason for the change without benefit of 2.
additional or supplemental information; s
A detailed description of the equipment, components, and processes involved and the interfaces 3.
with other plant systees; An evaluation of the change, which shows the predicted releases of radioactive materials in 4.
liquid and gaseous af fluents and/or quantity of solid waste that differ f ree those previously predicted in the license application and amendments thereto; f
An evaluation of the change, which shows the expected maximum suposures to member (s) of the 5.
the site boundary and to the general population that differ from those previously public at estimated in the license application and amendments thereto; A comparison of the predicted releases of radioactive materials, in liquid and gaseous af fluents 6.
and in solid waste, to the actual releases for the period prior to when the changes are to be f
made; An estimate of the exposure to plant operating personnel as a result of the change; and 7.
I Documentation of the fpet that the change was reviewed and found acceptable by PORC.
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Shall become of factive upon review and acceptance by PORC and approval by the Plant Manager.
3.
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- Licensee may choose to submit the information es11ed for in this Specification as part of the annual FSAR update.
222 Amendrent No.
83
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