ML20107M776

Proposed Tech Specs,Deleting Definitions Od Unrestricted Area, Restricted Area & Controlled Area

ML20107M776
Person / Time
Site:	Vermont Yankee
Issue date:	04/26/1996
From:	VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20107M774	List: ML20107M769 ML20107M776
References
NUDOCS 9605010117
Download: ML20107M776 (49)

Text

!

United States Nuclear Regulatory Commission VERMONT YANKEE NUCLEAR POWER CORPORATION April 19,19%

i Enclosure A Revised Technical Specification Pages l

i I

l i

l 9605010117 960426 PDR ADOCK 05000271 P PDR J

i l VYNPS I

1.0 DEFINITIONS X. Transition Boilino - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with )

j neither type being completely stable.

l 1 \

l Y. Surveillance Freauency - Unless otherwise stated in these specifications, periodic surveillance tests, checks, calibrations, l and examinations shall be performed within the specified surveillance

! intervals. These intervals may be adjusted plus 25%. The operating cycle interval is considered to be 18 months and the tolerance stated above is applicable.

Z. Surveillance Interval - The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable. These tests unless otherwise stated in these specifications may be waived when the instrument, component, or system is not required to be operable, but these tests shall be performed on the instrument, component, or system prior to being  !

required to be operable.

AA. Vital Fire Suppression Water System - The vital fire suppression water system is that part of the fire suppression system which protects those instruments, components, and systems required to perform a safe shecdown of the reactor. The vital fire suppression system includer che water supply, pumps, and distribution piping with ,

associated ser.tionalizing valves, wnich provide immediate coverage of the Reactor T,uilding, Control Roc.m Buildirc, and Diesel Generator Rooms.

BB. Source Check - The qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

CC. Dose Ecuivalent I -131 The dose equivalent I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131 I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion 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.

l EE. Deleted FF. Site Boundary - The site boundary is shown in Figure 2.2-5 in the FSAR.

GG. Deleted HH. Deleted Amendment No. M , G3, 43, %, 64, M4, W4, 4

VYNPS BASES: 3.2 (Cont'd) l setting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria. The instrumentation also -

covers the full range of spectrum breaks and meets the above criteria.

The high drywell pressure instrumentation is a backup to the water level instrumentation, and in addition to initiating ECCS, it causes isolation of Group 2, 3, and 4 isolation valves. For the complete circumferential break discussed above, this instrumentation will initiate ECCS operation at about the same time as the low-low water level instrumentation, thus, the results given above are applicable here also. Group 2 isolation i valves include the drywell vent, purge, and sump isolation valves. High i drywell pressure activates only these valves because high drywell pressure could occur as the result of nonsafety-related causes such as not purging the drywell air during startup. Total system isolation is j not desirable for these conditions and only the valves in Group 2 are i required to close. The water level instrumentation initiates protection j for the full spectrum of loss-of-coolant accidents and causes a trip of I all primary system isolation valves. j Venturis are provided in the main steam lir.er as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel

)i during a steam line break accident. In addition to monitoring steam )

flow, instrumentation is provided which causes a trip of Group 1 1 isolation valves. The primary function of the instrumentation is to I detect a break in the main steam line, thus only Group 1 valves are closed. For the worst case accident, main steam line break outside the ,

drywell, this trip setting of 140 percent of rated steam flow in 1 conjunction with the flow limiters and main steam line valve closure limit the mass inventory loss such that fuel is not uncovered, cladding l temperat2res remain less than 12950F and release of radioactivity to the environs is well below 10CFR100.

Temperature monitoring instrumentation is provided in the main steam line l tunnel to detect leaks in this area. Trips are provided on this j instrumentation and when exceeded cause closure of Group 1 isolation j valves. Its setting of ambient plus 950F is low enough to detect leaks l of the order of 5 to 10 gpm; thus, it is capable of covering the entire '

spectrum of breaks. For large breaks, it is a backup to high steam flow instrumentation discussed above, and for small breaks, with the resultant j small release of radioactivity, gives isolation before the limits of '

10CFR100 are exceeded.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure resulting from a control rod drop accident.

This instrumentation causes closure of Group 1 valves, the only valves required to close for this accident. With the established setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10CFR100 limits are not l exceeded for the control rod drop accident. With an alarm setting of 1.5 times normal background, the operator is alerted to possible gross fuel failure or abnormal fission product releases from failed fuel due to transient reactor operation.

Pressure instrumentation is provided which trips when main steam line pressure drops below 800 psig. A trip of this instrumentation results in closure of Group 1 isolation valves. In the refuel, shutdown, and startup modes, this trip function is provided when main steam line flow exceeds 40% of rated capacity. This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the control and/or bypass valves to open, resulting in a rapid depressurization and cooldown of the reactor vessel. The 800 psig trip Amendment No. &&, GB, 68, 64, 66, 76

VYNPS 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION 3.8 RADIOACTIVE EFFLUENTS 4.8 RADIOACTIVE EFFLUENTE Applicability: Applicability:

Applies to the release of all Applies to the required radioactive effluents from the surveillance of all radioactive plant. effluents released from the plant.

Obiective: Obiective:

To assure that radioactive To ascertain that all effluents are kept "as low as is radioactive effluents released reasonably achievable" in from the plant are kept "as low accordance with 10CFR50, as is reasonably achievable" in Appendix I and, in any event, accordance with 10CFR50, are within the dose limits for Appendix I and, in any event, Members of the Public specified are within the dose limits for in 10CFR20. Members of the Public specified in 10CFR20.

Specification: Specification A. Licuid Effluents: A. Licuid Effluents:

concentration concentration

1. The concentration of 1. Radioactive material in radioactive material in liquid waste shall be liquid effluents sampled and analyzed in released to Unrestricted accordance with Areas shall be limited requirements of to 10 times the Table 4.8.1. The concentrations specified results of the analyses in Appendix B to 10CFR shall be used in 1 Part 20.1001 - 20.2401, accordance with the Table 2, Column 2 for methods in the ODCM to radionuclides other than assure that the noble gases and concentrations at the 2x10 4 uCi/ml total point of release to activity concentration Unrestricted Areas are for all dissolved or limited to the values in entrained noble gases. Specification 3.8.A l.

2. With the concentration of radioactive material in liquid effluents released to Unrestricted Areas exceeding the limits of Specification 3.8.A.1, immediately take 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.

Amendment No. 83, 172

. . _ . . _ -. _ _ - - ~ .- . .. .

l VYNPS l

3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION l

B. Liouid Effluents: Dose B. Licuid Effluents: Dose

1. The dose or dose 1. Cumulative dose commitment to a member contributions shall be of the public from determined in accordance radioactive materials in with the methods in the liquid effluents ODCM at least once per l

released to Unrestricted month if releases during Areas shall be limited the period have to the following: occurred. I

a. During any calendar quarter:

less than or equal i to 1.5 mrem to the total body, and 1

less than or equal to 5 mrem to any organ, and

b. During any calendar year:

less than or equal to 3 mrem to the total body, and less than or equal to 10 mrem to any organ.

C. Liouid Radwaste Treatment C. Liouid Radwaste Treatment

1. The liquid radwaste 1. See Specification treatment system shall 4.8.B.l.

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 effluents released to Unrestricted Areas, when averaged with all other liquid l releases over the last month, would exceed 0.06 mrem to the total body, or 0.2 mrem to any organ.

l l

Amendment No. M, 173

VYNPS 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION L. Primary Containment L. Primary Containment

1. If the primary 1. The primary containment containment is to be shall be sampled prior Vented / Purged, it shall to venting / purging per be Vented / Purged through Table 4.8.2, and if the the Standby Gas results indicate '

Treatment System radioactivity levels in l whenever the airborne excess of the limits of I radioactivity levels in Specification 3.8.L.1, containment of the containment shall be Iodine-131, Iodine-133 aligned for or radionuclides in venting / purging through l particulate form with the Standby Gas half-lives greater than Treatment System. No 8 days exceed the levels sampling shall be specified in Appendix B required if the to 10CFR20.1001 - venting / purging is 20.2401, Table 1, through the Standby Gas Column 3. Treatment (SBGT) System.

2. With the requirements of Specification 3.8.L 1 not satisfied, immediately suspend all Venting / Purging of the containment.

3. During normal refueling and maintenance outages when primary containment is no longer required, then Specification 3.8.G shall supersede Specifications 3.8.L.1 and 2.

l M. Total Dose (40CFR190) M. Total Dose

1. The dose or dose 1. Cumulative dose commitment to a member contributions from of the public* in areas liquid and gaseous at and beyond the Site effluents shall be Boundary from all determined in accordance station sources is with Specifications limited to less than or 4.8.B.1, 4.8.F.1, and equal to 25 mrem to the 4.8.G.1.

total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem) over a calendar year.

• NOTE: For this Specification a menter of the public may be taken as a real individual accounting for his actual activities.

Amendment 'No. 83, 178

VYNPS TABLE 4.8.2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum Type of of Detection Gaseous Sampling Analysis Activity (LLD)

Release Type Frequency Frequency Analysis (uCi/ml)a A. Steam Jet Once per week Once per Xe-138, 1 x 10-4 Air Ejector Grab Sample week Xe-135, Xe-133, Kr-88, Kr-87, Kr-85M l B. Containment Prior to each Prior to Principal 1 x 10-9 (g)

Purge release / each Gamma Each Purge release / Emitters d '9 Grab Sample Each Purge and I-131 for Particu-lates C. Main Plant Once per Once per Principal 1 x 10-4 Stack monthC Grab monthC Gamma Sample Emitters d H-3 1 x 10-6 Continuous

• Once per I-131f 1 x 10-12 weekb Charcoal Sample Continuous

• Once per Principal 1 x 10-11 week b Gamma Particulate Emitters d '0 Sample and I-131 Continuous

• Once per Gross Alpha 1 x 10-11 month  ;

Composite l Particulate '

Sample Continuous

• Once per Sr-89, Sr-90 1 x 10-11 quarter Composite Particulate Sample Continuous Noble Gas Noble Gases 1 x 10-5 Monitor Gross Beta or Gamma Amendment No. M. 183

- . _ . . . . - .~. . - _ . - - . .- - - --

l l l r

VYNPS l

TABLE 4.8.2 NOTATION:

a. 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 /> after 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 thermal power change exceeding 25% of l rated thermal power in one hour, and analyses shall be completed within l 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.

l 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 I only if: (1) analysis shows that the dose equivalent I-131 concentration l in the primary coolant has increased more than a factor of 3 and the resultant concentration is at least 1 x 10-1 kCi/ml; and (2) the noble l 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: (a) analysis shows that the dose equivalent I-131 concentration in the primary coolant has not increased more than a factor of 3 and the resultant concentration is at least 1 x 10~1 pCi/ml; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

d. The principal gamma emitters for which the LLD specification will apply l are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, i Xe-133m, Xe-135 and Xe-138 for gaseous emissions, and Mn-54, Fe-59,  !

Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides l are to be detected and reported. Other peaks which are measurable and I identifiable, together with the above nuclides, shall also be identified I and reported. Nuclides which are below LLD for the analyses should not l be reported as being present at the LLD level for that nuclide, but as l "not detected". When unusual circumstances result in LLDs higher than '

required, the reasons shall be documented in the Annual Radioactive i Effluent Release Report. I

e. The ratio of the sample flow rate to the sampled stream flow rate shall l be known for the time period covered by each dose or dose rate I calculation made in accordance with Specifications 3.8.E.1, 3.3.F.1 and 3.8.G.1.

f. The gaseous waste sampling and analysis program does not explicitly require sampling and analysis at a specified LLD to determine the I-133 release. Estimates of I-133 releases shall be determined by counting the weekly charcoal sample for I-133 (as well as I-131) and assume a constant release rate for the release period.

g. Lower Limit of Detection (LLD) applies only to particulate form radionuclides identified in Table Notation d. above.

l l

l i

Amendment No. 84, 144, 184 l

l l

I VYNPS l

l BASES: 1 3.8 RADIOACTIVE EFFLUENTS 1

A. Liquid Effluents: Concentration I l This specification is provided to ensure that at any time the concentration of radioactive materials released in liquid waste effluents from the cite above background (Unrestricted Area for liquids is at the point of discharge from the plant discharge into  ;

Connecticut River) will not exceed 10 times the concentration levels specified in 10CFR Part 20.1001-20.2401, Appendix B, Table 2, Column 2. These requirements provide operational flexibility, compatible with considerations of health and safety, which may temporarily result in releases higher than the absolute value of the concentration numbers in Appendix B, but still within the annual average limitation of the Regulation. Compliance with the design objective doses of Section II.A of Appendix I to 10CFR Part 50 assure ,

that doses are maintained ALARA, and that annual concentration limits l of Appendix B to 10CFR20.1001-20.2401 will not be exceeded. I The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radionuclide and that an effluent concentration in air (submersion dose equal to 500 mrem /yr) was '

converted to an equivalent concentration in water. l 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 i Limiting Condition for Operation implements the guides set forth in i Section II.A of Appendix I. 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 material in liquid effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I, i.e., that conformance I 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, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in potable drinking water that are in excess of the requirements of 40CFR 141. No drinking water supplies drawn from the Connecticut River below the plant have been identified. The appropriate dose equations for implementation through requirements of the Specification are described in the Vermont Yankee Off-Site Dose Calculation Manual. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents were developed from the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", Revision 1, April 1977.

C. Licuid Radwaste Treatment The requirement that the appropriate portions of this system as indicated in the ODCM be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable" This specification inglements the requirements of 10CFR Part 50.36a and the design objective given in Section II.D of Appendix I to 10CFR Part 50. The Amendment No. 64, 185

I VYNPS l l

BASES: 3.8 (Cont'd) l specified limits governing the use of appropriate portions of the l liquid radwaste treatment system were specified as a suitable l

fraction of the dose design objectives set forth in Section II.A of '

Appendix I, 10CFR Part 50, for liquid effluents.

D. Liquid Holdup Tanks The tanks listed in this Specification include all outdoor tanks that contain radioactivity that are not surrounded by liners, dikes, or walls capable of holding the tank contents, or that do not have tank overflows and surrounding area drains connected to the liquid i radwaste treatment system. i l

Restricting the quantity of radi. active material contained in the specified tanks provides assurance that in the event of an l

uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10CFR Part 20.1001 -

20.2401, Appendix B, Table 2, Column 2, at the nearest potable water supply and in the nearest surface water supply in an Unrestricted Area.

E. Gaseous Effluents: Dose Rate The specified limits as determined by the methodology in the ODCM, restrict, at all times, the corresponding gamma and beta dose rates above background to a member of the public at or beyond the site boundary to (500) mrem / year to the total body or to (3,000) mrem / year to the skin. This instantaneous dose rate limit allows for 1 operational flexibility when off normal occurrences may temporarily I increase gaseous effluent release rates from the plant, while still i providing controls to ensure that licensee meets the dose objectives of Appendix I to 10CFR50. l 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 mrem / year for the highest impacted cow.

F. Gaseous Effluents: Dose from Noble Gases This specification is provided to implement the requirements of j Sections II.B, III.A, and IV.A of Appendix I, 10CFR Part 50. The  !

Limiting Condition for Operation implements the guides set forth in I Section II.B of Appendix I. The requirements provide operating l 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 material 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, i.e., that conformance I 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 public through appropriate pathways is unlikely to be substantially underestimated. The appropriate dose equations are l

l Amendment No. M , 186

VYNPS l

I BASES: 3.8 (Cont'd)

I. Ventilation Exhaust Treatment The requirement that the A0G Building and Radwaste Building HEPA filters be used when specified provides reasonable assurance that the l 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 I 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 NUREG-0473, Revision 2 (July 1979) as a suitable fraction of the guide set forth in Sections II.B and II.C of  ;

Appendix I, 10CFR Part 50, for gaseous effluents. j 4

J. Explosive Gas Mixture The hydrogen monitors are used to detect possible hydrogen buildups which could result in a possible hydrogen explosion. Automatic isolation of the off-gas flow would prevent the hydrogen explosion ,

and possible damage to the augmented off-gas system. Maintaining the I concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled.

K. Steam Jet Air Eiector (SJAEl Restricting the gross radioactivity release rate of gases from the main condenser SJAE provides reasonable assurance that the total body j 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 i this effluent is inadvertently discharged directly to the environment i without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10CFR Part 50.

L. Primary Containment (MARK I)

This specification provides reasonable assurance that releases from containment purging / venting operations will be filtered through the Standby Gas Treatment System (SBGT) so that the annual dose limits of 10CFR Part 20 for Members of the Public in areas at and beyond the Site Boundary will not be exceeded. The dose objectives of Specification 3.8.G restrict purge / venting operations when the Standby Gas Treatment System is not in use and gives reasonable assurance that all releases from the plant will be kept "as low as is reasonably achievable" The specification requires the use of SBGT only when Iodine-131, Iodine-133 or radionuclides in particulate form with half-lives greater than 8 days in containment exceeds the levels in Table 1, Colunn 3, to Appendix B of 10CFR 20.1001-20.2401 since the filter system is not considered effective in reducing noble gas radioactivity from gas streams.

l M. Total Dose (40CFR190)

This specification is provided to meet the dose limitations of 40CFR Part 190 to Members of the Public in areas at and beyond the Site Boundary. The specification requires the preparation and submittal i of a Specific Report whenever the calculated doses from plant i

radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly l 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 Amendment No. 84, 188

l l

l VYNPS BASES: 3.8 (Cont'd) l 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 l 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 corrected), in accordance with the provisions of 40CFR Part 190.11 l and 10CFR Part 20.2203(a)(4), 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 40CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10CFR Part 20. 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 subjects them to occupational exposures. For individuals in controlled areas who are considered Members of the Public per 10CFR20, the dose limits of 10CFR20.1301 apply since the licensee has the authority to control and limit access to these areas. ,

N. Solid Radioactive Waste This specification implements the requirements of 10CFR Part 50.36a with respect to the handling of solid radioactive waste (spent-resin and filter sludges only). The establishment and implementation of a Process Control Program (PCP), provides the operational guidelines by which proper dewatering of filter media and spent resins in preparation for off-site disposal is assured.

Amendment No. 83, 189

1 VYNPS TABLE 3.9 1 NOTATION s

l 1

NOTE 1 - With the number of channels operable less than required by-the '

minin um channels operable requirement, effluent releases may continue j provided that prier to initiating a release:

j a. At least two independent samples are analyzed in accordance with Specification 4.8.A.1, and I

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.

MOTE 2 - With the number of channels operable less than required by the minimum channels operable requirement, 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 l 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 i minimum channels operable requirement, effluent releases via this l 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. Camp performance curves may be used to estimate flow.

NOTE 4 - With the number of channels operable less than required by the minimum channels operable requirement, exert reasonable efforts 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 witi the methodology and parameters in the Off-Site Dose Calculation M&nual (ODCM). With a radioactive liquid ef fluent menitoring instrumer.tation channel alarm setpoint less conservative than a value which will ensure that the limits of l 3.8.A.1 are met during periods of release, immediately take action to suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable; or change the setpoint so it is acceptably conservative.

I 1

l l

Amendment No. 64, 194

T l

l l

l VYNPS TABLE 4.9.1 NOTATION (1) The Instrument Calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Institute for ,

I Standards and Technology) liquid radioactive source positioned in a j i reproducible geometry with respect to the sensor. These standards shall I

permit calibrating the system over its normal operating range of energy and rate. j (2) The Instrument Functional Test shall also demonstrate the Control Room l 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 controls 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 Off-Site Dose Calculation Manual (ODCM).

Amendment No. 84, 204

VYNPS TABLE 4.9.2 NOTATION (1) The Instrument Functional Test shall also demonstrate that automatic isolation of this pathway and 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 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:

(a) Instrument indicates measured levels above the alarm setpoint.

(b) Circuit failure.

(c) Instrument indicates a downscale failure.

(d) Instrument controls are not set in operate mode.

(3) The Instrument Calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Institute for Standards and Technology) radioactive source positioned in a reproducible geometry with respect to the sensor. These standards should permit

• calibrating the system over its normal operating range of rate capabilities.

(4) The Instrument Calibration shall include the use of standard gas samples (high range and low range) containing suitable concentrations, hydrogen balance nitrogen, for the detection range of interest per Specification 3.8.J.l. ,

i l

l l

l l

l Amendment No. 43, 206

l VYNPS BASES:

3.9 RADIOACTIVE EFFLUENT MONITORING SYSTEMS I 1

A. Liquid Effluent Instrumentation l l

The radioactive liquid effluent instrumentation is provided to l monitor and control, as applicable, the releases of radioactive materials in liquid effluents 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 10 times the concentration limits of Appendix B to 10CFR20.1001-20.2401, Table 2, Column 2, values.

Automatic isolation function is not provided on the liquid radwaste discharge line due to the infrequent nature of batch, discrete i volume, liquid discharges (on the order of once per year or less), I and the administrative controls provided to ensure that conservative i discharge flow rates / dilution flows are set such that the probability l of exceeding the above concentration limits are low, and the )

potential off-site dose consequences are also low. j B. Gaseous Effluent Instrumentation l

The radioactive gaseous effluent instrumentation is provided to l monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments 4 are provided to ensure that the alarm / trip will occur prior to l exceeding design bases dose rates identified in 3.8.E.1. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system.

C. Radiological Environmental Monitorino Procram The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of member (s) of the public resulting from the station operation. This monitoring program 1 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 effluent measurements and modeling of the environmental exposure pathways.

Ten years of plant operation, including the years prior to the implementation of the Augmented Off-Gas System, have amply demonstrated via routine effluent and environmental reports that plant effluent measurements and modeling of environmental pathways are adequately conservative. In all cases, environmar.tal 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

+.nuiro.imental monitoring program has therefore been significantly I modified as provided for by Regulatory Guides 4.3 (C.2.a) and 4.1 l (C.2.b), Revision 1, April 1975. Specifically, the air particulate {

and radiciodine air sampling periods have been increased to j semimonthly, based on plant effluent and environmental air sampling I data for the previous ten years of operation. An I-131 release rate j trigger value of 1 x 104 uCi/sec from i.he plant stack will require i that air sample collection be increased to weekly. The l

Amendment No. 63, 209

1 VYNPS 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 Figure 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 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 (B4 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 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. l C. Penetrations to the primary containment and piping passing l through such penetrations shall be designed in accordance with {

standards set forth in subsection 5.2 of the FSAR. l 5.5 Spent and New Fuel Storage 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 o gg of the fuel in the spent fuel storage pool shall be less than or equal to C.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.

I l

l l

i Amendment No. 14, G4, 143, 253 i

- .- .- . .~ -- . - . . .- -- . - -. - . _ , . ..

4 VYNPS i

8. . Process Control Program in-plant implementation.

9. Off-site Dose calculation Manual in-plant implementation.

J B. Radiation control standards and procedures shall be prepared, e

approved and maintained and made 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 radiation protection program shall be organized to meet the requirements of 10 CFR Part 20.

1. Paragraph 20.1601, " Control of Access to High Radiation Areas." In lieu of the " control device" or " alarm signal" required by Paragraph 20.1601(a), each high radiation area in which the intensity of radiation is greater than 100 mrem /hr at 30 cm, but less than 1000 mrem /hr at 30 cm, shall be barricaded and conspicuously posted as a high radiation area and entrance 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 following:

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.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.

c. A Health Physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose _ 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 frequency specified in the RWP.

The surveillance frequency will be established by the Plant Health Physicist.

The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than l 1000 mrem /hr at 30 cm, but less than 500 rad /hr at 1 meter.

In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift l Supervisor on duty and/or the Radiation Protection Manager.

• Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, providing they are following plant radiation protection procedures for entry into high radiation areas.

Amendment No. 36, 42, 43, 263

VYNPS 6.6 PLANT OPERATING RECORDS A. Records and/or logs relative to the following items shall be

, kept in a manner convenient for review and shall be retained i for at least five years:

i 1. Records of normal plant operation, including power i levels and periods of operation at each power level.

2. Records of principal maintenance activities, including inspection and repair or principal items of equipment pertaining to nuclear safety.

3. Records of reportable occurrences.

4. Records of periodic checks, inspection and/or calibrations performed to verify that surveillance requirements are being met.

5. Records of any special reactor test or experiments.

6. Records of changes made in the Operating Procedures.

l 7. Test results, in units of microcuries, for leak tests performed on licensed sealed sources.

l 8. Results of annual physical inventory verifying accountability of licensed sources on record.

B. Records and/or logs relative to the following items shall be i

recorded in a manner convenient for review and shall be retained for the life of the plant:

1. Records of substitution or replacement of principal items of equipment pertaining to nuclear safety.

2. Records of changes and drawing changes made to the plant as it is described in the Safety Analysis Report.

3. Records of plant radiation and contamination surveys.

4. Records of new and spent fuel inventory, transfers of

fuel, and assembly histories.

5. Records of radioactivity in liquid and gaseous wastes released to the environment.

6. Records of radiation exposure for all plant personnel, including all contractors and visitors to the plant for whom monitoring was required in accordance with 10 CFR 20.

i 7. Records of transient or operational cycling for those p; ant components that have been designed to 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.

4 1

Amendment No. -34 , 265

VYNPS i

10. Records for Environmental Qualification which are covered under the provisions of paragraph 6.9.

11. Records of analysis required by the Radiological Environmental Monitoring Program.

! l 12. Records of radioactive shipments.

6.7 REPORTING REOUIREMENTS 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

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 l involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the l 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 report.

l Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption of commencement of commercial power operation, or (3) 9 months following initial 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 shall be submitted at least every three months until all three events have been completed.

2. Annual Report An annual report covering the previous calendar year l shall be submitted prior to March 1 of each year. The l annual report shall include a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures i greater than 100 mrem /yr and their associated man-rem l 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.

l 1/ This tabulation supplements the requirements of 20.2206 of 10CFR Part 20.

Amendment No. 43, el, 266

VYNPS The dose assignment to various duty functions may be l estimates based on Self-Reading Dosimeter (SRD), 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 assigned to specific major work functions.

3. Monthly Statistical Report Routine 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 fifteenth 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.

4. Core Operatino Limits Report The core operating limits shall be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload cycle for the following: (a) The Average Planar Linear Heat Generation Rates (APLHGR) for Specifications 3.11.A and 3.6.G.la, (b) The Kg core flow adjustment factor for Specification 3.11.C., (c) The Minimum Critical Power Ratio (MCPR) for Specifications 3.11.C and 3.6.G.la, (d) The Linear Heat Generation Rates (LHGR) for Specifications 2.1.A.la, 2.1.B.1, and 3.11.B, and (e) The Power / Flow Exclusion Region for Specifications 3.6.J.la and 3.6.J.lb. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

Report, E. E. Pilat, " Methods for the Analysis of Boiling Water Reactors Lattice Physics," YAEC-1232, December 1980 (Approved by NRC SER, dated September 15, 1982).

Report, D. M. VerPlanck, " Methods for the Analysis of Boiling Water Reactors Steady State Core Physics,"

YAEC-1238, March 1981 (Approved by NRC, SER, dated September 15, 1982).

Report, J. M. Holzer, " Methods for the Analysis of Boiling Water Reactors Transient Core Physics,"

YAEC-1239P, August 1981 (Approved by NRC SER, dated September 15, 1982).

Report, S. P. Schultz and K. E. St. John, " Methods for the Analysis of Guide Fuel Rod Steady-State Thermal Effects (FROSSTEY); Code /Model Description Manual,"

YAEC-1249P, April 1981 (Approved by NRC SER, dated September 27, 1985).

Amendment No. 44, 44, 84, 95, 144, 146, 267

VYNPS (1) explanation of why gaseous radwaste was being discharged without treatment (Specification 3.8.H.1), or with resultant doses in excess of Specification 3.8.I.1, identification of any inoperable equipment or subsystems, and 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 provent a recurrence.

c. Total Dose, specification 3.8.M With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding the limits of Specification 3.8.M, prepare 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 3.8.M and includes the schedule for achieving conformance with these limits. This special report, required l by 10CFR Part 20.2203(a)(4), 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 radioactive material involved, and the cause of the exposure levels or concentrations. If 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 special report shall include a request for a variance in accordance with the provisions of 40CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

d. Radiolooical Environmental Monitorino, Specification 3.9.C With the level of radioactivity as the result of plant effluents in an environmental 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 from the receipt of the Laboratory Analyses a special report which includes an evaluation of any release conditions, environmental factors or other factors which caused the limits of Table 3.9.4 to be exceeded. This report is not required if the measured level of radioactivity was not the result of plant effluents, however, in such i an event, the condition shall be reported and i described in the annual Radiological Environmental Surveillance Report. l i

l i

Amendment No. 64, 95, 274 i

VYNPS 6.13 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

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 effluents for the purpose of demonstrating compliance with 10CFR50, Appendix I, 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 Annual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

l- a. Suf ficient information to support the change together with appropriate analyses or evaluations j justifying the change (s) and

b. A determination that the change will maintain the level of radioactive effluent control required by 10CFR20.1302, 40CFR190, 10CFR50.36a, and Appendix I to 10CFR Part 50, and not adversely impact the accuracy or reliability of effluent dose or setpoint calculations.

2. Shall become effective upon review by PORC and approved by the Manager of Operations (MOO).

3. Shall be submitted to the Commission in the form of a copy of the affected pages of the ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the magin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

6.14 MAJOR CHANGES TO RADIOACTIVE LIOUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS

• Licensee-initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):

1 A. Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the PORC. The discussion of each change shall contain:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10CFR Part 50.59

2. Sufficient detailed information to support the reason j for the change without benefit of additional or supplemental information;

, 3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;

• Licensee may choose to submit the information called for in this Specification as part of the annual FSAR update.

Amendment No. B4, GG, 404, 144, 278

United States Nuclear Regulatory Commission VERMONT YANKEE NUCLEAR POWER CORPORATION April 19,1996 Enclosure B Marked-Up Technical Specification Pages

VYNPS 1.0 DEFINITIONS X. Transition Boilina - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

Y. Surveillance Frequency - Unless otherwise stated in these specifications, periodic surveillance tests, checks, calibrations, and examinations shall be performed within the specified surveillance intervals. These intervals may be adjusted plus 25%. The operating cycle interval is considered to be 18 months and the tolerance stated above is applicable.

Z. Surveillance Interval - The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable. These tests unless otherwise stated in these specifications may be waived when the instrument, component, or system is not required to be operable, but these tests shall be performed on the instrument, component, or system prior to being required to be operable.

AA. Vital Fire Suppression Water System - The vital fire suppression water system is that part of the fire suppression system which protects those instruments, components, and systems required to perform a safe shutdown of the reactor. The vital fire suppression system includes the water supply, pumps, and distribution piping with associated sectionalizing valves, which provide immediate coverage of the Reactor Building, Control Room Building, and Diesel Generator Rooms.

BB. Source Check - The qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

CC. Dose Equivalent I-131 - The dose equivalent I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion 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_f,ilter_aludges. l gm

~,'

r- ,

,~ l y'( 3( p EE. Member (s V of the Public - Members of the public shaIl include all N s personsf'who are not/occupftionally agsociated with th,e' plant. This ;

category do 's not / inclu4e emp f \

/or v96 dors Als9 excl 8ed fr}oyees om th s ca6f t e utility, Jts cdntractorsgory  ; pie ps thefplant and p hoeptertp)esit to service e pme'nt of to ,

mak'e deli eries.yrsons_~ - v [ _ ._[

l_,__ .

- '/

v m _.

FF. Site Boundarv - The site boundary is shown in Figure 2.2-5 in the l

FSAR.

GG. Deleted HH. Deleted l

B4, 44, 74, 64, +44, 124 4 Amendment No. +9,

VYMPS BASES: 3.2 (Cont'd) setting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria. The instrumentation also covers the full range of spectrum breaks and meets the above criteria.

The high drywell pressure instrumentation is a backup to the water level instrumentation, and in addition to initiating ECCS, it causes isolation of Group 2, 3, and 4 isolation valves. For the complete circumferential break discussed above, this instrumentation will initiate ECCS operation at about the same time as the low-low water level instrumentation, thus, the results given above are applicable here also. Group 2 isolation valves include the drywell vent, purge, and sump isolation valves. High drywell pressure activates only these valves because high drywell pressure could occur as the result of nonsafety-related causes such as not purging the drywell air during startup. Total system isolation is not desirable for these conditions and only the valves in Group 2 are required to close. The water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip of all primary system isolation valves.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. In addition to monitoring steam flow, instrumentation is provided which causes a trip of Group 1 isolation valves. The primary function of the instrumentation is to detect a break in the main steam line, thus only Group 1 valves are closed. For the worst case accident, main steam line break outside the drywell, this trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limit the mass inventory loss such that fuel is not uncovered, cladding temperatures remain less than 12950F and release of radioactivity to the environs is well below 10CFR100.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in this area. Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves. Its setting of ambient plus 950F is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, it is a backup to high steam flow instrumentation discussed above, and for small breaks, with the resultant small release of radioactivity, gives isolation before the limits of 10CFR100 are exceeded.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure resulting from a control rod drop accident.

This instrumentation causes closure of Group 1 valves, the only valves required to close for this accident. With the established setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10CFR100_ limits _are_not . . .

exceeded for the control rod drop acciden f3y d,4DC K20 )( A V a6tC ygg McfsdeCf6ft6re16t3MMuR44mgdeq6tffp at$dn JWith an alarm setting of 1.5 times normal background, the opera or is a erted to possible gross fuel failure or abnormal fission product releases from failed fuel due to transient reactor operation.

Pressure instrumentation is provided which trips when main steam line pressure drops below 800 psig. A trip of this instrumentation results in closure of Group 1 isolation valves. In the refuel, shutdown, and startup modes, this trip function is provided when main steam line flow exceeds 40% of rated capacity. This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the control and/or bypass valves to open, resulting in a rapid depressurization and cooldown of the reactor vessel. The 800 psig trip Amendment No. GG, 66, 66, &4, 86 76

VYMPS 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION 3.8 RADIOACTIVE EFFLUENTS 4.8 RADIOACTIVE EFFLUEh,g, i Applicability: l Applicability- i Applies to the release of all Applies to the required radioactive effluents-from the surveillance of all radioactive l i

plant. effluents released from the plant.

Obiectives Obiective:

To assure that radioactive To ascertain that all effluents are kept 'as low as is reasonably achievable- in radioactive effluents released from the plant are kept 'as low accordance with 10CFR50, as is reasonably achievable" in Appendix I and, in any event, accordance with 10CFR50, are within the limits specified Appendix I and, in any event,

, in 10CFR20. . --

are within the limitspspecified

/ M n ot % Publ in 10CFR20.

Specificationi Specification ID hofs .

A. Liquid Eff:,uents: A. Liauid Effluents: '

Concent rat :,on concentration

~_

do s 1. The concentration of 1. Radioactive material in turesh.de -

radioactive material in liquid waste shall be

' < ru .s i id effluents I sampled and analyzed in

~~~~

re sa 4 t 'ftcp(AtMLe7 accordance with '

shall be-limited to%he requirements of Aggend;p 6 concentrations specified Table 4.8.1. The F P t AO, results of the analyses f' ED10CFE ED'f d , %b 44 shall be used in j 79, p3_7o, ggl/ 2 for accordance with the <

radion ides other than methods in th> ODCM to

( lkbk 2/olvem a noble gases and assure that the

\ -- ~'# 2x10~4 uCi/ml total- concentrations at the

~

activity concentration point ot releasev are for all dissolved or entrained noble gases.

limited to the values in '

Specification 3.8.A.1.

h sh44e/

2, With the concentration A/#AS of radioactive material in liquid effluents g releaaedQfelpr'Atwvpiq) exceeding tne limits of i

()(psf n,fec/ Specification 3.8.A.1 immediately take 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.

I Amendment No. 83 172

. _ _ ._. _ _ - . . . - - . - _ . . _ _ ~ . . _ . _ __. . ._ _ _ _.

VYNPS 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION B. f,iould Effluents: Dose B. Liould Effluents: Dose

! 1. The dose or dose 1. Cumulative dose commitment to a member contributions shall be of the public from determined in accordance radioactive materials in with the methods in the ip liquid s'fluents ODCM at least once per gf/f1g ,g)p) released frj>nr N month if releases during shall be 11m1ted to the the period have Off01 followings occurred.

a. Durang any calendar quarter:

less than or equal to 1.5 mrem to the total body, and less than or equal to 5 mrem to any organ, and

b. During any calendar years less than or equal to 3 mrom to the total body, and less than or equal to 10 mrem to any organ.

C. Liouid Radwaste Treatment C. Licuid Radwaste Treatment  :

1. The liquid radwaste 1. See specification treatment system shall 4.8.B.1.

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-Otl(05(d the liquid effluen g e en aver

+O w aT1 other liquid lVortsfdcNd releaseSover the last g ygg 3 j month, would exceed W 0.06 mrem to the total body, or 0.2 mrem to any organ.

i l

Amendment No. 83 173

1 l

i VYNPS <

l 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION

! L. Primary Containment L. Primary Containment 1

1. If the primary 1. The primary containment

' containment is to be shall be sampled prior

>f lodine-131' Iodine-133 Vented / Purged, it shall to venting / purging per be Vented / Purged through Table 4.8.2, and if the l or radionuclides m. the Standby Gas results indicate )

particulate form with half- I Treatment System radioactivity levels in l l lives greater than 8 days whenever the airborne excess of the limits of '

l exceed the lwels specified radi activi e-1 n Specification 3.8.L.1, m

. Appendix B to enneminmane e eed t e, the containment shall be 1,y,1, p, j f t d aligned for ii 10CFR20.lf 01-20.2401, 10CFR2 , pp nd , venting / purging through 1

'Q 1, Column 3. T ble/I, Co 1 ,d the Standby Gas ,

t 5 oceto. ~-

Treatment System. No sampling shall be

! 2. With the requirements of required if the l Specification 3.8.L.1 venting / purging is 4 l not satisfied, through the Standby Gas '

l immediately suspend all Treatment (SBGT) System.

Venting / Purging of the l containment. i

3. During normal refueling and maintenance outages when primary containment is no longer required, then Specification 3.8.G ,

(

shall supersede l Specifications 3.8.L.1 and 2.

M. Total Dosel(40Ch N M. Total Dose

! 1. The dose or dose 1. Cumulative dose i commitment to a member contributions from <

of the public*ffrom all liquid and gaseous

'l0 offoj qf station sources is effluents shall be limited to less than or detenmined in accordance and be$ood equal to 25 mrem to the with Specifications

]Lc $il t total body or any organ (except the thyroid, 4.8.B.1, 4.8.F.1, and 4.8.G.I.

which is limited to less than or equal to 75 mrem) over a calendar year.

I l

• NOTE: For this Specification a member of the public may be taken as a real

individual accounting for his actual activities.

Amendment No. 83 178 I

VYNPS TABLE 4.8.2 l RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM l Lower Limit Minimm Type of of Detection Gaseous Sampling Analysis Activity (LLD) ,

Release Type Frequency _ . _

Frequency Analysis (uCi/ml)a i l A. Steam Jet once per week once per Xe-138, 1 x 10-4

! Air Ejector Grab Sample week Xe-135, Xe-133, Kr-88, Kr-87, Kr-85M l B. Containment Prior to each Prior to -

Principal 1 x 10 h i Purge each relea ef )camma Each Purge Emittershi (f T _.

Sind T-130 C. Main Pla Once ger Once ger Principal 1 x 10-4 Stack month Grab month Gamma Sample Emittersd l H-3 1 x 10-6 i Continuous

• Once per I-131 f 1 x 10-12 {

week l Charcoal Sample Continuous

• Once per Principal 1 x 10-11 weekb g ,,,

d Particulate Sample Emitters /h h

Continuous

• Once per GrossAlpha\ 1 x 10-11 -

month Composite Particulate Sample Continuous

• Once per Sr-89, Sr-90f 1 x 10-11 quarter Composite Particulate

[ Sample Continuous Noble Gas Noble Gases' 1 x 10-5 Monitor GrossBetaf or Gamma

/ [Oc $1 Por s

$('o b Sarrfk (t kd- f

. Pa(lic uld e s l

2 1

I Amendment No. 83 183

VYNPS TABLE 4.8.2 NOTATION:

a. 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 /> after 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 thermal 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 I-131 concentration in the primary coolant has increased more than a factor of 3 and the resultant concentration is at least 1 x IQ-1 pCi/ml; 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 (a) analysis shows that the dose equivalent I-131 concentration in the primary coolant has not increased more than a factor of 3 and the resultant concentration is at least 1 x 10-1 pCi/ml; and (2) the noble gas monitor shows that offluent activity has not increased more than a factor of 3.

d. The principal gansna emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emissions, and Mn-F4, Fe-59, Co-58, Co-60, En-65, Mo-99, Cs-134, Cs-137, co-141 and Ce-l u for particulate emissions. This list does not mean that only thene nuclides are to be detected and reported. Other peaks which are meanrable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below LLD for the analyses should not be reported as being present at the LLD level for that nuclide, fiut as "not detected". When unusual circumstances result in LLDs higher than l required, the reasons shall be documented in the Annual Radioactive Effluent Release Report.

e. The atio of the sample flow rate to the sampled stream flow rate shall be knc.wn 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...

f. The gaaeous waste sampling and analysis program does not explicitly requira sampling and analysis at a specified LLD to determine the I-133 releare. Estimates of I-133 releases shall be determined by counting the weekly charcoal sample for I-133 (as well as I-131) and assume a consrant roles.se rate for the release period. ._

~

. ~ . . _ . _ . , _ _ _.

D C, lCh)O lrW $ D k2 Chm

• l} f culoh h

fuI '

((;(hOpuclicSP1 YC4 YIP 5 D k ek

• O NCN . ,,/

.___ ~ --- ~ _

x -__

w_ _ _ __ _ . _ .-- -

l 1

l Amendment No. 84,144 184

1 VYNPS BASES:

hhb Y Y _

3.8 RADIOACTIVE EFFLUEffrS pp A. Liquid Effluents: Concentration h

This specification is p vided to ensure that the concentration of radioactive materials r leased in liquid waste effluents from the site above background t the point of discharge _fronuthe_plW ingharge_into_Connecti_ cut River) willJbe les's than' the ,

glevptsspe'ifiedin10pFRPart c 20, A Table I ncentration 3 i

ts lijnitati fi prov des additional'ppendi/'B, Column 2/

teri a assurtnce t)(at t levels off diopetive in ystposup withi in )hdies pi wate.r' outsjde th ite idill fesul (1) the Sec fon II A design obj iv of ApprendlyI, 10C Par 0, t a me r of tfie pub c, (2 the ',

limits of 10CF Part 0.106 (e) to he pgp,ulatio .f M e g ]4 O)Thethat concentration Xe-135 is thelimit for nobleradionuclide controlling gases is based upon the assumption andf&ivwPtfin air 1 O (submez mM was converted to an equivalenLg.pncentration in water, u 1 atiptiaFCI55tiTdF16nion Raiiio1ogiedIMTFc~ETFn3f 6P3)

~

j f f'2** f MaTL e Uta , F 5ter&aAQ B. Licuid Effluents: Dose i This specification is provided to implement the requirements of l 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 I. 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 material in liquid effluents will be kept 'as low as is reasonably l achievable". The Surveillance Requirements implement the j requirements 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 an individual through appropriate pathways is unlikely to be substantially underestimated. In addition, there is* reasonable assurance that the operation of the facility will not result in radionuclide concentrations in potable drinking water that are in l l

excess of the requirements of 40CFR 141. No drinking water supplies drawn from the Connecticut River below the plant have been identified. The appropriate dose equations for implementation through requirements of the Specification are described in the Vermont Yankee Off-Site Dose Calculation Manual. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents were developed from the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", Revision 1, April 197,.

C. Liquid Radwaste Treatment t

' The requirement that the appropriate portions of this system as indicated in the ODCM be used, when specified, provides assurance that the releases of radioactive materials in liquid 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 governing the use of appropriate portions of the Amendment No. 83 185

Page 185 Insert:

l A

not exceed 10 times the concentration levels specified in 10CFR Part 20.1001-20.2401, Appendix B, Table 2, Column 2. These requirements provide operational flexibility, compatible with considerations of health and safety, which may temporarily result in releases higher than the absolute value of the concentration numbers in Appendix B, but still within the annual average limitation of the Regulation. Compliance with the design objective doses of Section II.A of Appendix ! to 10CFR Part 50 assure that doses are maintained ALARA, and that annual concentration limits of Appendix B to 10CFR20.1001-20.2401 will not be exceeded.

l l

VYNPS 1A,S1S A S - 3.8 (Cont'd) i liquid radwaste treatment system were specified as a suitable l fraction of the dose design objectives set forth in Section II.A of  !

Appendix I, 10CFR Part 50, for liquid effluents.

l D. Liauid Holduo Tanks l

.The tanks listed in this Specification include all outdoor tanks that contain radioactivity that are not surrounded by liners, dikes, or ,

walls capable of holding the tank contents, or that do not have tank  !

overflows and surrounding area drains connected to the liquid "

radwaste treatment system.

Restricting the quantity of radioactive material contained in the l' l0cFR 'Pcs4 specified tanks provides assurance that in the event of an b' ')AD /I uncontrolled release of the tanks' contents, the resultina oncentr tions_Would be_.l g than the limits of)1OcT1VPrn(V2 %

1. 9 Nmr d / at the nearest portable water supply

~

' [F ndW b k 9 g h 70lv#,1 f /@g~ face water supply in an hrestricted an 'Ylis Hearest sur ea.

C E. Gaseous Effluents: Dose Rate

~

[TEsf fpectffcat (n Y vided do ensure'that t5(U o's7 at7r ime at' and' bey'on'd the} site)cundary/from gaseous effl'uents )w hin Il be ,

the annual d(se lipits of OCFR Part' 20. 'phe annual dose , limits ar'e i the 4ssociatied wi the cone'entratiphs of IfCFR Part 20, / l

- ~

/Appe, doses, ndix'B, Table II, olumn 1./ These limits p his mstantaneous .

assurance that! radioactive material dis' charged ycvide re'asonablein gaseous offfuents/i dose rate limit allows for operational

'/will'not result the within or'outside' in'the exposure of member (s)/of the site'boundaryI to annual aver',public either/ age contentrations/r flexibility when off / exceeding Part 20 10the CFRlimits Partspecified 20'.106(b(iVAppendix B, Table II of'10CER /. 1 normal occurrences at times (be'within the site boundary,fFor member (s)/of the the occupancy j of publi f

                                                                                                                                    'who,dua'lmay' the indiv.i I may temporarily                                                                                                                   f                       ,

increase gaseous will be suf ficiently diffusionlow tofcompensate _ factor _above that for any increkse for'the_s.ite_ in the[/[The boundary effluent release rates atmospheri;c.iinits specifIed l as determined by the methodology in the ODCM, restrict, at all times, the corresponding gamma and beta dose rates ( from the plant, w.hile above background to a member of-the public at or beyond the site still providing i boundary to (500) mrom/ year to the total body or to :(3,000) mrem / year controls to ensure tha j theskinj iicensee meets the dose objectives of 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 Appendix 1 t thyroid dose rate above background to an infant via the L10CFR50. cow-milk-infant pathway to 1500 mrem / year for the GiWppf/pejrDdpt f M* td irnpaciecI CM F. Gaseous Effluentst Dose from Noble Gases This specification is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I, 10CFR Part 50. The Limiting Condition for operation implements the guides set forth in Section II.B of Appendix I. 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 material 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, 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 public through appropriate pathways is unlikely to be 4 substantially underestimated. The appropriate dose equations are l Amendment No. 83 186

VYNPS BASES: 3.8 (Cont'd) l I. Ventilation Exhaust Treatment 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 I 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 NUREG-0473, Revision 2 (July 1979) as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I, 10CFR Part 50, for gaseous effluents. J. Explosive Gas Mixture  ! The hydrogen monitors are used to detect possible hydrogen buildups which could result in a possible hydrogen explosion. Automatic isolation of the off-gas flow would prevent the hydrogen explosion and possible damage to the augmented off-gas system. Maintaining the concentration of hydrogen below its flammability limit provides i assurance that the releases of radioactive materials will be controlled. i l K. Steam Jet Air Eiector (SJAEL l Restricting the gross radioactivity release rate of gases from the main condenser SJAE provides reasonable assurance that the total body l 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 specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10CFR Part 50. L. Primary Containment (MARK I) (ggGT) fu Nemkr5 This specificatlon provides r asonable assurance that releases from containment purging / venting perations will be filtered through the ffhe dbhcin Standb-KYY~2 Gas Treatment System so that the annual dose limits of 10CFR atethe f te A undary will not be exceeded. The dose GNAS objectives of Specification 3.8.G restrict purge / venting operations I when the Standby Gas Treatment System is not in use and gives I and reasonable assurance that all releases from the plant will be kept l q egnd "as low as is reasonably achievable". M. Total Dose (4Wd/90) l his sp ification is provided to meet the dose limitations of 40CFR h dus Of Part 1_ f-mmh The 7f5co) on re@ ires' t~he~ preparation specificati,ygrap4KpCFKf#1Wortg2LV gygg and IGL40*ksubmittal of a Specific Report whenever the calculated doses from 4 La70 O plant radioactive effluents exceed twice the design objective doses Opanc(kr Mordi ht of Appendix I. For sites containing up to 4 reactors, it is highly J$h Gov /dage ' unlikely that the resultant dose to a member of the public will N - -_x exceed the dose limits of 40CFR Part 190 if the individual reactors remain within the reporting requirement level. The Special Report I will describe a course of action that should result in the limitation i 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 1 Amendment No. 83 188

Page 188 Insert: B The specification requires the use of SBGT only when lodine-131, lodine-133 or radionuclides in particulate form with half-lives greater than 8 days in containment exceeds the levels in Table 1, Column 3, to Appenoix 8 of 10CFR20.1001-20.2401 since the filter system is not considered effective in reducing noble gas radiocctivity from gas streams. 1 l l 1 l l l l l i l

I VYNPS ( BASES: 3.8 (Cont'd) bZD)((k l corrected), in acco ance with the provisions of OCFR Part : 90.11 l and 10CFR Part 20. , fulfillstherequir'e~)entsof40CFRPartis m considered 190 unt to be1 NRC a ti staff ely requestjand kction is l completed. The variance only relates to the li its of 40CFR I l Part 190, and does not apply in any way to the ther require nents for i dose limitation of 10CFR Part 20, as addressed in Specificat ion 3.8.A and 3.8.E. Anindividualisnotconsideredapemberofthe(p)blic l l during any period in which he/she is engaged in carrying out any operation that 7W, art php rpet')eas pepyy N. Solid Radioactive Waste l This specification implements the requirements of 10CFR Part 50.36a with respect to the handling of solid radioactive waste (spent resin and filter sludges only). The establishment and implementation of a Process Control Program (PCP), provides the operational guidelines by which proper dewatering of filter media and spent resins in preparation for off-site disposal is assured. l l

          ., .          . _ _ ~           _ _        _

_ - -- -~ l subjects them to occupational exposures. For individuals in controlled areas who are considered Members of the Public per 10CFR20, the dose limits of 10CFR20.1301 apply since the licensee } has the authority to control and limit access to these areas. (

        ' - ~ ~ ~                                                                                                  

_ .__ _ _ _ . _ _ _ _ _ _ ~ _ _ __ - l l i 4 Amendment No. 83 189

t VYNPS TABLE 3.9.1 NOTATION NOTE 1 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases may continue provided that prior to initiating ~a releases

a. 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, effluent releases via this pathway may continue provided that, at least once per 24 hours, grab samplesarecollectedandanalyzedforgrossradioacgivity(betaor gamma) at a lower limit of detection of at least 10- microc'urie/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 hours during actual releases. pump performance curves may l be used to estimate flow. I i NOTE 4 - With the number of channels operable less than required by the minimum channels operable requirement, exert reasonable efforts 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 and parameters in the l off-site Dose calculation Manual (ODcM). With a radioactive liquid j effluent monitoring instrumentation channel alarm setpoint less f.8.A.1 are met during periods of release, immediately take action to 3(conservativethanavaluewhichwillensurethatthelimitsof I suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable; or change the j setpoint so it is acceptably conservative. l l i l l l l Amendment No. 83 194 i

VYNPS

                                                             ~~

n. TABLE 4.9.1 NOTATION AIh[hh h [kMpfe/J g,p/

                                                                ~-

[r2

                                               -.-~ ~

(1) The shallInstrument Calibration for radioactivity include theJan_ofAknown (traceablemeasurement instr.umeqtation] tLo NationalgatVaty fahshirA#) liquid radioactive source positioned in a reproducible ometry 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 controls 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 Off-Site Dose Calculation Manual (ODCM). i i l l Amendment No. 83 204

VYNPS TABLE 4.9.2 NOTATION (1) The Instrument Functional Test shall also demonstrate that automatic isolation of this pathway and 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 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: (a) Instrument indicates measured levels above the alarm setpoint. (b) Circuit failure. [ . (c) Instrument indicates a downscale failure. g/qd,g/g ano/ (d) Instrument controls are not set in operate mode. I#ff (3) The Instrument Calibration for radioactivity measurement _ instrumentation shall include the use of a known (traceable to National]B9tp44 4 ((((h ~i radioactive source positioned in a reproducible geometry with re oc to the sensor. These standards should permit calibrating the system over its normal operating range of rate capabilities. (4) The Instrument Calibration shall include the use of standard gas samples l (high range and low range) containing suitable concentrations, hydrogen  ! balance nittogen, for the detection range of interest per Specification 3.8.J.1. I l Amendment No. 83 206

VYNPS i BASES: 3.9 RADIOACTIVE EFFLUENT MONITORING SYSTEMS _ A. Liquid Effluent Instrumentation The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm setpoints for these instruments are to ensure that the alarm will occur priog to exceediny (EG'7nETliErg*,3 LL0d$AW2]>. /0 6m1 7% remvw/nb AMfr et//)wx4, 8 7C

                /6tWro./$d/ ~c?d.MS/ 756] 2,(9/a,pn 2 [4r/dja Automatic isolation functio,n is not provided o,n the liquid radwaste discharge line due to the infrequent 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@W)GE~tfl#j [ concentration limits are low, and the l potential off-site dose cons quenc.aLAre also low. l G H ~ B. Gaseous Effluent Instrumentation gggg The radioactive gaseous effluent inp rumenEation is provided 86.I monitor and control, as applicable,p the releases of radioactive materials in gaseous effluents d t'ing actual or potential releases of gaseous effluents. The alarm / p setpoints for these instruments are provided to ensure that th ala / trip will occur prior to exceeding @t44fo#gRyPptg/J . This instrumentation also includes provisions for monitoring and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. C. Radiolocical Environmental Monitorino Procram The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of member (s) of the public resulting from the station operation. This monitoring program l implements Section 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 effluent measurements and modeling of the environmental exposure pathways. Ten years of plant operation, including the years prior to the implementation of the Augmented Off-Gas System, have amply demonstrated via routine effluent and environmontal reports that plant effluent measurements and modeling of environmental pathways are adequately conservative. In all cases, 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 Regulatory Guides 4.3 (C.2.a) and 4.1 (C.2.b), Revision 1, April 1975. Specifically, the air particulate and radiciodine air sampling periods have been increased to semimonthly, based on plant effluent and environmental air sampling data for the previous teg years of operation. An I-131 release rate uCi/sec from the plant stack will require trigger value of 1 x 10~ that air sample collection be increased to weekly. The Amendment No. 83 209

VYNPS 5.0 DeSiaN FEATURES OS 4)e// Gs arms df km clpr M CM 70 5.1 Site # ### # # sqfe g - The station is located on the property on the west bank of the l Connecticut River in the Town of ernon, Vermont, which the < Vermont Yankee Nuclear Power Cor ration either owns or to which I it has perpetual rights and ease nts. The site plan showing the exclusion area boundary, bound for gaseous effluents @ y boundary for liquid effluentg on Figure 2.2-5 in the M AR. The minimum distance to the boundary of the exclusion area as defined in 10CFR100.3 ks 910 feet. No part of the site shall be s. eld or leased and no structure shall be located on the site except 2tructures owned by the Vermont , Yankee Nuclear Power Corporatiin or related utility companies and used in conjunction with norma utility operations. J 5.2 Reactor A. The core shall consist of not more than 368 fuel assemblies. B. The reactor core shall etntain 89 cruciform-shaped control rods. The control mater.a1 shall be boron carbide powder (B4 c) or hafnium, or a e>mbination 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 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. 1 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 (K of the fuel when dry effective multiplication factoris less than 0.90 is and when flooded,gg)less than 0.95. B. The K ,g of the fuel in the spent fuel storage pool shall be less Inan 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. 44, 84, 123 253

l VYNPS

8. Process Control Program in-plant-implementation.

9. Off-Site Dose Calculation Manual in-plant I implementation. j B. Radiation control standards and procedures shall be prepared, approved and maintained and made available to all~ station '

personnel. These procedures shall show permissible radiation I exposure, and shall be consistent with the requirements of 10 CFR Part 20. This radiation protection program shall be organized to meet the require nts of 10 CFR Part 20.

                        - [              Paya ra
                                      - (ga
                                                     ~ . tLB7[
                                                  " f fn lieu kah didna'. Qh6 /si Mgtf M the " control device",or " alarm ragraph $44071d{27" each highL@/(,0/[o c70, l(#01 "M4d d j

signal" required by radiation area in w ch the intensity of radiation is OCCFf 5 ID lh h facltdrw OhW pr f shall be barricaded and D conspicuously posse as a high radiation area and 3 p yrf/M entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.* Any individual or group of individuals permitted to enter such areas shall (ajer Nao MMUN {/ be provided with one or more of the following: Ol 30tg d kiSihad a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

     %0twelhr0A30W                      b. A radiation monitoring device which continuously           A
      \~

integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. , Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them,

c. A Health Physics qualified individual (i.e.,

qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the i activities within the area and who will perform t periodic radiation surveillance at the frequency specified in the RWP. The surveillance frequency will be established by the Plant Health Physicist. The above procedure sha leo apply to each high radiation area in which t intensity of radiction is gf jo (,nI bd l00 greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorize.d entry into Moo 90md/hr,/ such areas and the keys shall be maitamined under the administrative control of the Shi t Supervisor on duty (d}Wj6 and/or the {Ippc N e piWJf?fg .

                                                                'Icbil7 i

• Health Physics personnel shall be exempt from the RWP issuance requirement j during the perforinance of their assigned radiation protection duties, j providing they are following plant radiation protection procedures for entry I into high radiation arean.

Amendment No. M , 44, 83 263

VYNPS 6.6 PLANT OPERATING RECORDS ! A. Records and/or logs relative to the following items shall be kept in a manner convenient for review and shall be retained for at least five years: l 1. Records of normal plant operation, including power i levels and periods of operation at each power level.

2. Records of principal maintenance activities, including inspection and repair or principal items of equipment

pertaining to nuclear safety.

3. Records of reportable occurrences.

l 4. Records of periodic checks, inspection and/or calibrations performed to verify that surveillance requirements are being met.

5. Records of any special reactor test or experiments.

6. Records of changes made in the Operating Procedures.

                                   /NW?'WWhhY&

F] ff. Test results, in units of microcuries, for leak tests performed on licensed sealed sources. l Aff. U Results of annual physical inventory verifying accountability of licensed sources on record. B. Records and/or logs relative to the following items shall be recorded in a manner convenient for review and shall be retained for the life of the plant: l 1. Records of substitution or replacement of principal l items of equipment pertaining to nuclear safety. l l 2. Records of changes and drawing changes made to the plant ! as it is described in the safety Analysis Report. t

3. Records of plant radiation and contamination surveys.

4. Records of new and spent fuel inventory, transfers of fuel, and assembly histories.

5. Records of radioactivity in liquid and gaseous wastes

       -                              released to the environment.

[ 6. Records of radiation exposure for all plant personnel, b including all contractors and visitors to the plant 'n 1D r W 8M ! #gg', h g accordance with 10 CFR 20. hgg3 (pqj7 7. Records of transient or operational cycling for those l U plant components that have been designed to operate l safely for a limited number of transients or operational , cycles.  !

8. Records of inservice inspections of the reactor coolant system.

4 9. Minutes of meetings of the Plant Operation Review 4 Committee and the Nuclear Safety Audit and Review Board. Amendment No. 31 265

VYNPS

10. ' Records for Environmental Qualification which are covered under the provisions of paragraph 6.9. ,

11.

                  ^~

Records of analysis required by the Radiological 1 Monitoring _ Program;

                     \ L ,. vironmentY RCords 0       tadloocWP       SMpH    s~)

6.7 REPORTING REOUIREMENTS-~ - - ~ 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

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 different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the 'i 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 report.

Startup reports shall be submitted within (1) 90 days l following completion of the startup test program, (2) 90 days following resumption of commencement of commercial power operation, or (3) 9 months following initial 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 consnercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

2. Annual Report An annual report covering the previous calendar year shall be submitted prior to March 1 of each year. The annual report shall include a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem 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. cNO& . 1/ Thistabulationsupplementstherequirementsof20.Q$$of10CFRPart 20. Amendment No. eg, 83 266

                     -        .-.                   -   ~      . . - -      ._

i VYNPS -- ggD) SeI(hab** The dose assignm vari duty unctions may be estimates based o (k4 - simete , 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 assigned to specific major work functions.

3. Monthly Statistical Report Routine reports of operating statistics and shutdown I experience shall be submitted on a monthly basis to the 'l Office of Management Information and Program control, U.S.  ;

Nuclear Regulatory Commission, Washington, D.C. 20555, with I a copy to the appropriate Regional Office, to arrive no later than the fifteenth 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.

4. Core Operatino Limits Report The core operating limits shall be established and l documented in the Core Operating Limits-Report (COLR) before I each reload cycle or any remaining part of a reload cycle for the followings (a) The Average Planar Linear Heat Generation Rates (APLHCR) for Specifications 3.11.A and 3.6.G.la, (b) The Kg core flow adjustment factor for Specification 3.11.C., (c) The Minimum Critical Power Ratio (MCPR) for Specifications 3.11.C and 3.6.G.la, (d) The Linear Heat Generation Rates (LHGR) for Specifications 2.1.A.la, 2.1.B.1, and 3.11.B, and (e) The Power / Flow Exclusion Region for Specifications 3.6.J.la and 3.6.J.lb. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

Report, E. E. Pilat, ' Methods for the Analysis of Boiling l Water Reactors Lattice Physics," YAEC-1232, December 1980 (Approved by NRC SER, dated September 15, 1982), l Report, D. M. VerPlanck, " Methods for the Analysis of Boiling Water Reactors Stsady State Core Physics," YAEC-1238, March 1981 (Approved by NRC, SER, dated September 15, 1982). Report, J. M. Holzer, " Methods for the Analysis of Boiling Water Reactors Transient Core Physics," YAEC-1239P, August 1981 (Approved by NRC SER, dated September 15, 1982). Report, S. P. Schultz and K. E. St. John, " Methods for the Analysis of Guide Puel Rod Steady-State Thermal Effects (FROSSTEY): Code /Model Description Manual," YAEC-1249P, April 1981 (Approved by NRC SER, dated September 27, 1985). Amendment No. 44, 44, 64, 96, 444,146 267

1 VYNPS (1) explanation of why gaseous radwaste was being F discharged without treatment (Specification 3.8.H.1), or with resultant doses in excess of Specification 3.8.I.1, identification of any inoperable equipment or subsysteam, and 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.

c. Total Dose, Specification 3.8.M With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding the limits of Specification 3.8.M, prepare 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 3.8.M and includes the schedule for achieving conformance th d...es) mi_m Thisshallspecial report, required c70.o7728 3 h by 10CFR Pa , include an analysis.

that estimates the radiation exposure (dose) to a j member of the public from station sources, including all effluent pathways and direct ) radiation, for the calendar year that includes the j release (s) covered by this report. It shall also  : describe levels of radiation and concentrations of l radioactive material involved, and the cause of the exposure levels or concentrations. If 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 special report shall include a request for a variance in accordance with the provisions of 40CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

d. Radiolocical Environmental Monitorina, specification 3.9.C with the level of radioactivity as the result of plant effluents in an environmental 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 from the receipt of the Laboratory Analyses a special report which includes an evaluation of any release conditions, environmental factors or other factors which caused the limits of Table 3.9.4 to be exceeded. This report is not required if the measured level of radioactivity 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.

Amendment No. 84, 95 274 l

[ ' N .~_ % VYNPS hr &f&b bWW$4h] 4 occ kA Ut IOCFS 50l fA (%dA t 6.13 OFF-SITE DOSE CALCULATION MANUAL (ODCM) F l An Off-site Dose calculation Manual shall contain the current methodology and parameters used in the calculation o ff-site doses due to radioactive gaseous and liquid effluent in the calculation of gaseous and liquid offluent monitoring alarm / trip i setpoints, and in the conduct of the environmental radiological i monitoring program. 1 1 A. Licensee initiated changes to the ODCH:

1. Shall be sulunitted to the Commission in the Annual Radioactive Effluent Release Report for the period in l which the change (s) was made effective. This submittal shall contain:

                                  - - 'N x_      _-                              a. Sufficien M 41444)information to support the i maintain the level of radioactive                       N* *A' " ' * " ""g change  1 a inf out t

ne n. o-efiluent control required by

                                                                                                                                /

8 10CFR20.1302 40CFR190- of e pag of he DCM ic wer e ng 1 . 10CFR50.36a, and Appendix ! to I e ch en har d = or ide wit the r vi

                                                                       , Together with a'ppropriate anal 10CFR Part 50, and not adversely impact the accuracy or reliability of              eva untions justifying the change (s) aAd l

effluent dose, or setpoint Adeterminationthatthechanawild

                                                                                                                          ~

l b.

                                                                                                                            ~~
                        ~, . ,_
                                          , l                          ,_

y , Documentation of the' fact that the change has been { [,[ fev4eGod by'PORC,and appro ed by'the Managerfo'f / )

                                                                                                                  /- / //

s N ~ ~p'O rations _j .(MOOT . f- ,/ f' Md 2. Shall become effective upon review by PORC and approved by the Manager of Operations (MOO). l 6.14 M UOR CHANGES TO RADIOACTIVE LIOUID. CASEOUS, AND SOLID WASTE TREATMENT SYSTEMS

• Licensee-initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):

l A. Shall be reported to the conunission in the Annual Radioactive Effluent Release Report for the period in which the , evaluation was reviewed by the PORC. The discussion of each 1 i I change shall contains l l l 1. A summary of the evaluation that led to the I determination that the change could be made in  ! accordance with 10CFR Part 50.59; l

2. Sufficient detailed information to support the reason for the change without benefit of additional or supplemental information;

3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant

systems; i

)

• Licensee may choose to submit the information called for in this specification as part of the annual FSAR update.

Amendment No. M, 96, -1.M 144 278

m._ - . . . _ - . _ _ . . _ _ . _ _ . . _ _ _ . _ _ . _ . . _ . . _ _ . . _ _ _ . _ _ . Page 278 Insert: i C i

3. Shall be submitted to the Commission in the form of a copy of the affected pages of tite ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g. month / year) the change was implemented.

i I l i l - l l 1 I i e _}}