ML20217N300
| ML20217N300 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 10/21/1999 |
| From: | VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20217N298 | List: |
| References | |
| NUDOCS 9910280170 | |
| Download: ML20217N300 (11) | |
Text
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VERMONT YANKEE NUCLEAR POWER CORPORATION Docket No. 50-271
_BVY 99-134 i.
I Attachment 3 Vermont Yankee Nuclear Power Station i
Proposed Technical Specification Change No. 228 Administrative Change l
Marked-up Version of the Current Technical Specifications ]
1 9910290170 991021.
r
., ,, VYNPS
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1.0 DEFINITIONS
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' factors used for this calculation shall be those listed in NRC Regulatory Guide 1.109, Revision 1, October 1977.
DD. Solidification - Solidification shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. Suitable forms include dewatered resins and filter sludges.
EE. Deleted FF.SiteBoundary-Thesiteboundary'isshownin([ly... 2. 2 ; 2.. u.m r;A51 GG, Deleted pp .y HH. Deleted II. Off-Site Dose Calculation Manual (ODCM) - A manual containing the current methodology and parameters used in the calcui3 tion of off-site doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduction of the environmental radiological monitoring program.
JJ. Process Control Program (PCP) - A process control program shall contain the sampling, analysis, tests, and determinations by which wet radioactive waste from liquid systems is assured to be converted to a form suitable for off-site disposal.
KK. Gaseous Radwaste Treatment System - The Augmented Off-Gas System (AOG) is the gaseous radwaste treatment system which has been designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for" delay or holdup for the purpose of reducing the total radioactivity prior to release to the
]
l environment.
LL. Ventilation Exhaust Treatment System - The Radwaste Building and AOG Building ventilation HEPA filters are ventilation exhaust treatment systems which have been designed and installed to reduce radioactive material in particulate form in gaseous effluents by passing ventilation air through HEPA filters for the purpose of removing radioactive particulates from the gaseous exhaust stream prior to release to the environment. Engineered safety feature atmospheric cleanup systems, such as the Standby Gas Treatment (SBGT) System, are not considered to be ventilation exhaust treatment system components.
MM. Vent / Purging - Vent / Purging is the controlled process of discharging air or gas from the primary containment to control temperature, pressure, humidity, concentration or other operating conditions.
NN. Core Operating Limits Report - The Core Operating Limits Report is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.C.
Plant operation within these operating limits is addressed in individual j specifications.
~ i Amendment No. 84, 4Mr M4, MB, 171 5 l
I VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION
- 3. The indicated core flow 3. The surveillance is the sum of the flow requirements of 4.6.F.1 indication from each of and 4.6.F.2 do not apply the twenty jet pumps. to the idle loop and If flow indication associated jet pumps failure occurs for two when in single loop or more jet pumps, operation.
immediate corrective action shall be taken. 4. The baseline data If flow indication for required to evaluate the all but one jet pump conditions in cannot be obtained Specifications 4.6.F.1 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> an and 4.6.F.2 shall be orderly shutdown shall acquired each operating be initiated and the cycle. Baseline data reactor shall be in a for evaluating 4.6.F.2 cold shutdown condition while in single loop within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. operation shall be updated as soon as G. Sincle Loop operation practical after entering single loop operation.
- 1. The reactor may be started and operated or operation may continue with a single recirculation loop provided that:
I
- a. The designated l adjustments for !
APRM flux scram and rod block trip settings (Specifi-cations 2.1.A.l.a )
and 2.1.B.1, I Table 3.1.1 and Table 3.2.5), rod )
block monitor trip setting (Table 3.2.5), MCPR fuel cladding integrity safety limit (Specifi-cation 1.1.A), and MCPR cperating I limits and MAPLHGR limits, provided in l the Core Operating l Limits Report, are initiated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, either these adjustments must be completed or the reactor brought to Hot Shutdown.
/
( Detector Le s A and C of ne LPRM strin per core oct t plus de ctor Levels A d C of one LP string in th center of t core shal be monitor .
/
Amendment No. H, 94, H4, H+,146 122
VYNPS 71 , . 3.8. LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS
' OPERATION
- 2. With the calculated dose 2. Cumulative dose from the release of contributions from radioactive materials in direct radiation from liquid or gaseous plant sources shall be effluents exceeding determined in accordance twice the limits of with the methods in the Specifications ODCM. This requirement 3.8.3.1.a. 3.8.B.1.b, is applicable only under 3.8.F.1.a, 3.8.F.1.b, conditions set forth in.
3.8.c.1.a, or 3. 8.G .1.b, Specification 3.8.M.2.
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.
N. Solid Radioactive Waste N. Solid Radioactive Waste
- 1. The solid radwaste 1. Verification of system shall be used in solidification of wet accordance with a waste shall be performed Process control Program as required and in j accordance with the (9wet dradioactive re-t*^=
rrrihf r 1?f i o process waste Process Control Program.
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(spent resins / filter sludges) to meet I shipping 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.
I 1
1 I
Amendment No. 83 179 c
_ _ _ .._ _ _ _ _ _ . = .
4
., J s VYNPS
- 5. 0- DESIGN FEATURES
{ 5.1 Jite, The station is located on the property on the west bank of the Connecticut River in the Town of Vernon, Vermont, which the Vermont Yankee Nuclear Power Corporation either owns or to which it has perpetual rights and eas===nts. The site plan showing the exclusion area boundary, boundary for gaseous effluents, boundary for liquid effluents, as well as areas defind ner 10CFR20 as " controlled areas
- and "unre:stricted areas
- are on@i. a 2.24 e L 26 The min 4== distance to the boundary 10CFR100.3 is 910 feet, tne exclusion area as defined in gg o g.a No part of the site shall be sold or leased and no structure shall be located on the site except structures owned by the Vermont Yankee Nuclear Power Corporation or related utility companies and used in conjunction with normal utility operations.
5.2 Reactor A. The core shall consist of not more than 368 fuel assemblies.
B. The reactor core shall contain 89 cruciform-shaped control rods.
The control material shall be boron carbide powder (B 4c) or hafnium, or a combination of the two.
5.3 Reactor Vessel The reactor vessel shall be as described in Table 4.2-3 of the FSAR.
The applicable design codes shall be as described in subsection 4.2 4
of the FSAR.
5.4 Containrnent 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.
5.5 Spent and New Fuel Storace A. The new fuel storage facility shall be such that the effective multiplication factor (K,gg) of the fuel when dry is less than 0.90 and when flooded is less than 0.95.
B. The K,gg of the fuel in the spent fuel storage pool shall be less than or equal to 0.95.
C. Spent fuel storage racks may be moved (only) in accordance with written procedures which ensure that no rack modules are moved over fuel assemblies.
Amendment No. M, M. us. 151 253
-VERMONT YANKEE NUCLEAR POWER CORPORATION
' ' Docket No. 50-271 BVY 99-134 Attachment 4
. Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 228 Administrative Change Retyped Technical Specification Pages
1
,- .~. I VERMONT YANKLt NUCLEAR POWER CORPORATION
. BYY 99-1J4 / Attachment 4 / Page 1
' Listing of Affected Technical Specifications Panes Replace the Vermont Yankee Nuclear Power Station Technical Specifications pages listed below with the revised pages. The revised pages contain venical lines in the margin indicating the areas of change.
Remove . Insert 5 5 l
122 122
'179 179 253 253 1
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l l
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VYNPS 1.0 DEFINITIONS factors used for this calculation shall be those listed in NRC J Regulatory Guide 1.109, Revision 1, October 1977.
DD. Solidification - Solidification shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.
Suitable forms include dewatered resins and filter sludges.
EE. seleted FF. Site Boundary - The site boundary is shown in plant drawing 5920-6245.
GG. Deleted HH. Deleted I
II. Off-Site Dose Calculation Manual (ODCM) - A manual containing the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduction of the environmental radiological monitoring program. j JJ. Process Control Program (PCP) - A process control prog sm shall contain the sampling, analysis, tests, and determinations by which wet radioactive waste from liquid systems is assured to be converted to a form suitable for off-site disposal.
KK. Gaseous Radwaste Treatment System - The Augmented Off-Gas System (AOG) is the gaseous radwaste treatment system which has been designed and I installed to reduce radio &ctive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
LL. Ventilation Exhaust Treatment System - The Radwaste Building and AOG Building ventilation HEPA filters are ventilation exhaust treatment !
systems which have been designed and installed to reduce radioactive material in particulate form in gaseous effluents by passing ventilation air through HEPA filters for the purpose of removing radioactive particulates from the gaseous exhaust stream prior to release to the environment. Engineered safety feature atmospheric cleanup systems, such as the Standby Gas Treatment (SBGT) System, are not considered to be ventilation exhaust treatment system components.
MM. Vent / Purging - Vent / Purging is the controlled process of discharging air or gas from the primary containment to control temperature, pressure, humidity, concentration or other operating conditions.
NN. Core Operating Limits Report - The Core Operating Limits Report is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.C. Plant operation within these operating limits is addressed in individual specifications.
Amendment No. e4, 4Mr M1, MB, -94 5
VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OFERATION
- 3. The indicated core flow 3. The surveillance is the sum of the flow requirements of 4.6.F.1 indication from each of and 4.6.F.2 do not apply the twenty jet pumps. If to the idle loop and flow indication failure associated jet pumps when occurs for two or more in single loop operation.
jet pumps, immediate corrective action shall 4. The baseline data be taken. If fl~' required to evaluate the indication for all but conditions in one jet pump cannot be Specifications 4.6.F.1 obtained within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 4.6.F.2 shall be an orderly shutdown shall acquired each operating be initiated and the cycle. Baseline data for reactor shall be in a evaluating 4.6.F.2 while cold shutdown condition in single loop operation within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, shall be updated as soon as practical after G. Sipgle Loop Operation entering single loop operation.
- 1. The reactor may be started and operated or operation may continue with a single recirculation loop provided that:
- a. The designated adjustments for APRM flux scram and rod block trip settings (Specifi-cations 2.1.A.1.a and 2.1.B.1, Table 3.1.1 and Table 3.2.5),
rod block monitor trip setting (Teble 3.2.5), MCPR fuel cladding integrity safety limit (Specifi-cation 1.1.A), and MCPR operating limits and MAPLHGR limits, provided in the Core Operating Limits Report, are initiated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, either these adjustments must be crapleted or the reactor brought to Hot Shutdown.
Amendment No. 8, 44, M4, 444, 444 122
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VYNPS I 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION
- 2. With the calculated dose 2. Cumulative dose from the release of contributions from direct
-radioactive materials in radiation from plant liquid or gaseous sources shall be effluents exceediaq twice determined in accordance the limits of with the methods in the Specifications 3.8.B.1.a, ODCM. This requirement 3.8.B.1.b, 3.8.F.1.a, is applicable only under i 3.8.F.1.b, 3.8.G.1.a, or conditions set forth in 3.8.G 1.b, calculations Specification 3.8.M.2.
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.
N. Solid Radioactive Waste N. Solid Radioactive Waste
- 1. The solid radwaste system shall be used in 1. Verification of accordance with a Process solidification of wet Control Program to waste shall be performed !
process wet radioactive as required and in waste (spent accordance with the r3 sins / filter sludges) to Process Control Program.
meet shipping and burial j ground requirements. '
- 2. With the provisions of Specification 3.8.N.1 not ,
satisfied, suspend l shipments of defectively processed or defectively packaged solidified wet radioactive wastes from the site.
Amendment No. 43 179
l VYNPS l' . .
5.0 DESIGN FEATURES
! 5.1 Site The station is located on the property on the west bank of the Connecticut River in the Town of Vernon, Vermont, which the Vermont Yankee Nuclear Power Corporation either owns or to which it has perpetual rights and easements. The site plan showing the exclusion area boundary, boundary for gaseous effluents, boundary for liquid effluents, as well as areas defined per 10CFR20 as " controlled areas" and " unrestricted areas" are on plant drawing 5920-6245. The minimum distance to the boundary of the exclusion area as defined in
>CFR100.3 is 910 feet.
No part of the site shall be sold or leased and no structure shall be located on the site except structures owned by the Vermont Yankee Nuclear Power Corporation or related utility companies and used in conjunction with normal utility operations.
5.2 Reactor A. The core shall consist of not more than 368 fue; assemblies.
l B. The reactor core shall contain 89 cruciform-shaped control rods. The control material shall be boron carbide powder (B.C) or hafnium, or a combination of the two.
5.3 Reactor Vessel The reactor vessel shall be as described in Table 4.2-3 of the FSAR.
The applicable design codes shall be as described in subsection 4.2 of the FSAR.
, 5.4 Containment 1
A. The principal design parameters and applicable design codes for j 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 i through such penetrations shall be designed in accordance with standards set forth in subsection 5.2 of the PSAR.
5.5 Spent and New Fuel Storage A. The new fuel storage facility sha. .e such that the effective multiplication factor ( K.rt ) of the el when dry is less than 0.90 and when flooded is less than 0.95.
B. The Karr of the fuel in the spent fuel storage pool shall be less than or equal to 0.95.
C. Spent fuel storage racks may be moved (onl 4.n accordance with written procedures which ensure that no r$y) ck modules are moved over fuel assemblies.
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Amendment No. 44, G3, 433, 4&4 253 I I