ML20141D201
| ML20141D201 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/19/1997 |
| From: | Milano P NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20141D207 | List: |
| References | |
| NUDOCS 9706270008 | |
| Download: ML20141D201 (25) | |
Text
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20566-0001
The Nuclear Regulatory Commission (the Comission) has found that:
l A.
The application for amendment by Vermont Yankee Nuclear Power Carporation (the licensee), dated August 27, 1993, as supplemented November 9, 1993, April 26, 1996, and September 25, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; i
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical l
Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-28 is hereby l
amended to read as follows:
(B) Technical Soecifications l
The Technical Specifications contained in Appendix A, as revised through Amendment No. 151, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
i 9706270008 970619 PDR ADOCK 05000271 P
o o
. l 3.
This license amendment is effective as of the date of its issuance, to be implemented within 60 days of issuance.
1 FOR THE NUCLEAR REGULATORY COMMISSION
~
-c-Patrick D. Milano, Acting Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
i Changes to the Technical i
Specifications Date of Issuance: June 19, 1997 J
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i ATTACHMENT TO LICENSE AMENDMENT NO.151 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 i
Replace the following pages of Appendix A, Technical Specifications, with the attached pages as indicated. The revised pages are identified by amendment l
number and contain vertical lines indicating the areas of change.
i Remove Insert 4
4 76 76 172 172 l
173 173 178 178 183 183 l
184 184 185 185 l
186 186 188 188 189 189 194 194 204 204 206 206 209 209 253 253 263 263 265 265 266 266 267 267 274 274 278 278
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VY14PS 1.0 DEFINITIONS X.
Transition Boilino - Transition boiling means the boiling regime l
between nucleate and film boiling. Transition boiling is the regime l
in which both nucleate and film boiling occur intermittently with neither type being completely stable.
l l
Y.
Surveillance Frecuency - Unless otherwise stated in these l
specifications, periodic surveillance tests, checks, calibrations, and examinations shall be performed within the specified surveillance l
intervals. These intervals may be adjusted plus 25%.
The operating l
cycle interval is considered to be 18 months and the tolerance stated l
above is applicable.
Z.
Surveillance Interval - The surveillance interval is the calendar I
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 l
system is not required to be operable, but these tests shall be performed on the instrument, component, or system prior to being l
required to be operable.
L AA. Vital Fire Suppression Water System - The vital fire suppression l
water system is that part of theefire 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 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 Regulatoky Guide 1.109, Revision 1, October 1977.
DD. Solidification - Solidification shall be the conversion of wet wastes l
into a form that meets shipping and burial ground requirements.
i l
Suitable forms include dewatered resins and filter sludges.
l EE.
Deleted FF. Site Boundarv - The site boundary is shown in Figure 2.2-5 in the FSAR.
GG.
Deleted HH.
Deleted I
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L l
l 4
Amendment No. H M, M, M,'M, M, W, 151
e s
VYNPS 1
l 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 raeets the above criteria.
The high drywell pressure instrumentation is a backup to the water level l
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 l
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 i
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 i
limit the mass inventory loss such that fuel is not uncovered, cladding i
temperatures remain less than 1295 F 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 95'F 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.
Por 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 l
exceede.d 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 i
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 depresgurization and cooldown of the reactor vessel.
The 800 psig trip Amendment No. BG, 64, 64, 64, 66, 151 76
_.m g
VYNPS 4
l 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION 3.8 RADIOACTIVE EFFLUENTS 4.8 RADIOACTIVE EFFLUENTS Aeolicability:
Aeolicability:
i i
Applies to the release of all Applies to the required radioactive effluents from the surveillance of all radioactive j
plant.
effluents released from the q
plant.
1 j
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 f
Appendix I and, in any event, accordance with 10CFR50, 4
are within the dose limits for Appendix I and, in any event, Hembers of the Public specified are within the dose limits for in 10CFR20.
Members of the Public specified in 10CFR20.
Soecification:
Soecification 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 i
Part 20.1001 - 20.2401, accordance with the Table 2, Column 2 for methods in the ODCM to radionuclides other than assure that the i
noble gases and concentrations at the 2x10-4 udi/ml total point of relaase 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 i
concentration to within the above limits.
j f
Amendment No. 64, 151 172 1
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v 1
8 1
VYNPS l
1 J
3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS l-OPERATION l
j B.
Liouid Effluents:
Dose B.
Licuid Effluents:
Dose 1
j 1.
The dose or dose 1.
Cumulative dose
^
commitment to a member contributions shall be i
of the public from determined in accordance 4
radioactive materials-in with the methods in the liquid effluents ODCM at least onca per released to Unrestricted month if releases during Areas shall be limited the period have to the following:
occurred.
a.
During any calendar quarter:
l less than or equal j
to 1.5 mrem to the i,
total body, and h
less than or equal to 5 mrem to any organ, and l
b.
During any calendar years j
less than or equal to 3 mrem to'the total body,-and i
3 less than or equal i
j to 10 mrem to any organ.
j C.
Licuid Radwaste Treatment C.
Licuid Radwaste Treatment i
1.
The liquid radwaste 1.
See Specification' j
treatment system shall 4.8.B.1.
q be used in its designed modes of bperation 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.
r Amendment No. 64,151 173
l I
VYNPS 1
3.8 LIMITING CONDITIONS FOR l 4.8 SURVEILLANCE REQUIREMENTS OPERATION L.
I-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 whenever the airborne excess of the limits of 1
radioactivity levels in Specification 3.8.L.1, l
containment of the containment shall be Iodine-131, Iodine-133 aligned for or radionuclides in venting / purging through particulate form with the Standby Gas l
half-lives greater than Treatment System.
No J
l 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 i
Column 3.
Treatment (ScGT) 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.
i
- NOTE:
Fdr this Specification a member of tha public may be taken as a real f
individual accounting for his actuc1 activities.
1 Amendment No. &&, 151 178 l
a VYNPS 1
TABLE 4.8.2 RADIOACTIVF GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM j
i Lower Limit j
Minimum Type of of Detection Gaseous Sampling Analysis Activity (LLD)
Release Type Frequency Frequency Analysis (uCi/ml)a j
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, i
Kr- ~8 7,
\\
Kr-85M l
C. Containment Prior to each Prior to Principal 1 x 10-9 (g)
Durge release /
each Gamma Each Purge release /
Emitters
'9 I
d Grab Sample Fach Purge and 1-131 for Particu-lates C. Main Plant once per Once per, Principal 1 x 10-4 Stack monthC Grab monthC Gamma d
Sample Emitters H-3 1 x 10-6 i
I Continuous
- Once per I-131f 1 x 10-12
'l b
week Charcoal Sample Continuous
- Once per Principal 1 x 10-11 weekb Gamma d
Particulate Emitters
'S Sample and I-131 Conginuous' Once per Gross Alpha 1 x 10-11 l
month
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Composite 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 l
i Amendment No. M, 151 183
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 i
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 nay 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 10-1 pCi/ml; and (2) the noble gas monitor shows that effluent activity has increased more than a factor of 3.
s 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 coclant 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 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 are to be detected and reported.
Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below LLD for the analyses should not be reported as being present at the LLD level for that nuclide, but as "not detected".
When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Annual Radioactive Effluent Release Report.
4 e.
The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.8.E.1, 3.8.F.1 and 3.8.G.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 radionu~clides identified in Table Notation d. above.
i Amendment No. 63, 144,151 184
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VYNPS j
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BASES:
3.8 RADIOACTIVE EFFLUENTS A.
Licuid Effluents:
Concentration l
This specification is provided to ensure that at any time the concentration of radioactive materials released in liquid waste effluents from the site 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 specif.ed in 10CFR Part 20.1001-20.2401, Appendix 3, Table 2, Colunn 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 d
l that doses are maintained ALARA, and that annual concentration limits of Appendix B to 10CFR20.1001-20.2401 will not be exceeded.
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.
/
B.
Licuid 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 l
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 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 thdt 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 l
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 l
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.
l C.
Licuid Radwaste Treatment l
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 l
be kept "as low as is reasonably achievable". This specification i
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 l
[
Amendment No. 64,151 185
)
VYNPS BASES:
3.8 (Cont'd) specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10CFR Part 50, for liquid effluents.
D.
Licuid Holdup Tanks a
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 specified tanks provides assurance that in the event of an 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 operational flexibility when off normal occurrences may temporarily increase gaseous effluent release rates from the plant, while still providing controls to ensure that licensee meets the dose objectives of Appendix I to 10CFR50.
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 Eff1 dents:
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 substantially underestimated. The appropriate dose equations are 7
Amendment No. M, 151 186
VYNPS BASES:
3.8 (Cont'd)
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 assurance that the releases of radioactive materials will be controlled.
K.
Steam Jet Air Eiector (SJAE)
Restricting the gross radioactivity release rate of gases from the main condenser SJAE provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This 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, Column 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 of a Specific Report whenever the calculated doses from plant 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. 64,161 188
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 i
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.
i r
1 1
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Amendment No. 64, 151 189
_._ -. _ _ _ -. _ _.~._._. _ _ _ _ _ _ _. -.m.
m_ _ _ _ _.
._-_.m..
VYNPS' j.
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 release:
i l
a.
At least two independent samples are analyzed in accordar.ce with Specification 4.8.A.1, and b.
At least two technically qualified members of the Facility Staff i
independently verify the release rate calculations and discharge line valving.
y a
l 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, grab samples are collected and analyzed for gross radioactivity (beta or
,[
gamma) at a lower limit of detection of at least 10-7 microcurie /ml.
NOTE 3 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.
Pump performance curves may be used to estimate flow.
NOTE 4 - With the number of channels operable less than required by the minimum channels operable requirement,-exert reasonable effortsEto 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 Off-Site Dose Calculation Manual (ODCM). With a radioactive liquid effluent monitoring-instrumentation 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.
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Amendment No. 64, 151 194
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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 Standards and Technology) liquid radioactive source positioned in a reproducible geometry with respect to the sensor. These standards shall permit calibrating the system over its normal operating range of energy and rate.
(2) The Instrument Functional Test shall also demonstrate the Control Room alarm annunciation occurs if any of the following conditions exists:
(a) Instrument indicate measured levels above the alarm setpoint.
(b) Circuit failure.
(c) Instrument indicates a downscale failure.
(d) Instrument 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).
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Amendment No. 63, 151 204
l a
VYNPS l
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:
4 (a) Instrument indicate measured levels above the alarm setpoint.
]
i (b) Circuit failure.
l (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.
I I
(b) Circuit failure.
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(c) Instrument indicates a downscale failure.
l l
(d) Instrument controls are not set in operate mode.
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(3) The Instrument Calibration for radioactivity weasurement instrumentation j
shall include the use of a known (traceable to National Institute for l
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 l
balance nitrogen, for the detection range of interest per Specification 3.8.J.l.
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Amendment No. 64,151 206
VYNPS BASES:
3.9 RADIOACTIVE EFFLUENT MONITORING SYSTEMS A.
Licuid 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 i
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 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 i
l of exceeding the above concentration limits are low, and the potential off-site dose consequences are also low, B.
Gaseous Effluent Instrumentation "he radioactive gaseous effluent instrumentation is provided to 3
monitor and control, as applicabl, the releases of radioactive 9
materials in gaseous ef fluer.ts during actual or potential releases of i
gaseous effluents.
The alarm / trip setpoints for these instruments are provided to ensure that the alarm / trip will occur prior to j
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.
l C.
Radiolooical Environmental Monitorino Procram i
l 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 implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby 3
supplements the radiological effluent monitoring program by verifying l
that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the j
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, 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 Pequlatory 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 ten years of operation.
An I-131 release rate 4
trigger value of 1 x 10 uCi/sec from the plant stack will require that air sample collection be increased to weekly.
The Amendment No. 63,151 209
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VYNPS 5.0 DESIGN FEATURES 5.1 Site j
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" i
l 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 i
10CFR100.3 is 910 feet.
j No part of the site shall be sold or leased and no structure shall be i
located on the site except structures owned by the Vermont Yankee Nuclear Power Corporation or related utility companies and used in l
conjunction with normal utility operations.
5.2 Reactor A.
The core shall consist of not more than 368 fuel assemblies.
B.
The reactor core shall contain 89 cruciform-shaped control rods.
The control material shall be boron carbide powder (B C) or 4
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.; of the FSAR.
B.
The secondary containment shall be as described in subsection 5.3 i
of the FSAR and the applicable codes shall be as described in l
Section I2.0 of the FSAR.
l C.
Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in subsectica 5.2 of the FSAR.
5.5 Spent and New Fuel Storace A.
The new fuel storage facility shall be such that the effective multiplication factor 0.90 and when flooded is cg) of the fuel when dry is less than (K
less than 0.95.
B.
The K,gg of the fuel in the spent fuel storage pool shall be less than or equal to 0.95.
C.
Spent fuel storage racks may be moved (only) in accordance with written procedures which ensure that no rack modules are moved over fuel assemblies.
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f Amendment No. 34, 63, tea, 151 253 I
VYNPS B.
Process Control Program in-plant implementation.
9.
Off-Site Dose Calculation Manual in-plant implementation.
B.
Radiation' control standards and procedures shall be prepared, approved and maintained and made available to all station personnel. These procedules 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 i
meet the requirements of 10 CFR Part 20.
l 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 I
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.
t The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater tha,n l
1000 mrem /hr at 30 cm, but less than 500 rad /hr at 1 meter.
In addition, locked doors shall be provided ta 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, j
. providing they are following plant radiation protection procedures for entry into high radiation areas.
Amendment No. 36, 44, &&, 151 263
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VYNPS l
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:
1.
Records of nornal plant operation, including power levels and periods of operation at each power level.
2.
Records of principal maintenance activities, including l
inspection and repair or principal items of equipment pertaining to nuclear safety.
3.
Records of reportable occurrences.
I 4.
Records of periodic checks, inspection and/or l
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 ph'ysical 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:
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 wast'es 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.
7.
Records of transient or operational cycling for those plant components that have been designed to operate safely for a limited number of transients or operational cycles.
8.
Records of insarvice inopections of the reactor coolant system.
9.
Minutes of meetings of the Plant Operation Review Committee and the Nuclear Safety Audit and Review Board.
Amendment No.
&+,
151 265
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k VYNPS l
10.
Records for Environmental Qualification which are r
l covered under the provisions of paragraph 6.9.
11.
Records of analysis required by the Radiological Environmentel Monitoring Program.
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12.
Records of radioactive shipments.
6.7 REPORTING REOUIREMENTS l
In addition to the applicable reporting requirements of Title 10 l
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.
Startuo 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 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.
I Startup repbrts shall be submitted within (1) 90 days following completion of the startup test program, b(2) 90 days following resumption of commencement of i
ommercial power operation, or (3) 9 months following i
initial criticality, whichever is earliest.
If the l
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 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, if e.g.,
reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.
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l J/ This tabulation supplements the requirements of 20.2206 of 10CFR Part 20.
Amendment No. 4G, 64, 151 266
VYNPS The dose assignment to various duty functions may be j
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 sha?.1 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 Operatina 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 L
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 i
Region for Specifications 3.6.J.la and 3.6.J.1b.
The i
analytical methods used to determine the core operating lf.mits 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. VarPlanck, " Methods for the Analysis o'f Boiling Water Reactors Steady State Core Physics,"
YAEC-1238, Ma.ch 1981 (Approved by NRC, SER, dated September 15, 1982).
l 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).
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Amendment No. 44, 44, 64. 96, 444, 444,151 267
l VYNPS (1) explanation of why gaseous radwaste was being discharged without treatment (Specification 3.8.H.1), or with resultan'c 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 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 i
i defines the corrective action (s) to be taken to l
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 i
exposure levels or concentrations.
If the estimated doses exceed any of the limits of i
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 l
include a request for a variance in accordance with l
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.
1 I
d.
Radioloaical Environmental Monitorino, Specification 3.9.C With the level of radioactivity as the result of plant effluents in an environmental samplir.g 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 i
Surveillance Report.
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274 Amendment No. 64, 95,151 I
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l VYNPS
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6.13 OFF-SITE DOSE CALCULATION MANUAL (ODCML 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.
l A.
Licensee initiated changes to the ODCM:
i 1.
Shall be submitted to the Commission in the Annual l
Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:
l a.
Sufficient information to support the change together with appropriate analyses or evaluations 1
l justifying the change (s) and b.
A determination that the change will maintain the level of radioactive effluent control required by 1
10CFR20.1302, 40CFR190, 10CFR50.36a, and Appendix I I
to 10CFR Part 50, and not adversely impact the accuracy or relfability of affluent dose or setpoint calculatient 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 legible 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 syst' ems (liquid, gaseous, and solid):
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 for the change without benefit of additional or supplemental information; j
3.
A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems; 1
l
- Licensee may choose to submit the information called for in this Specification as part of the annual FSAR update.
Amendment No. &&, 96, tG3, 144,161 278