ML20099K431

Proposed Tech Spec Making Administrative Changes Re Radiological Effluent Tech Specs

ML20099K431
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	03/04/1985
From:	VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20099K416	List: ML20099K415 ML20099K431
References
NUDOCS 8503200269
Download: ML20099K431 (20)

Text

m TABLE OF CONTENTS GMERAL Paste No'.

1.0 DEFINITIONS..................................................................................

1 5.0 DESIGN FEATURES..............................................................................

188 6.0 ADMINISTRATIVE CONTROLS......................................................................

190 6.1 ORGANIZATION.................................................................................

190 6.2 REVIEW AND AUDIT.............................................................................

194 A. Plant Ope ra tion Review Cotienit te e..........................................................

194 B. Nu c lea r Sa fe ty Aud it and Review Contenittee.................................................

196 6.3 ACTION TO BE TAKEN IN THE EVENT OF AN ABNORMAL OCCURRENCE IN PLANT OPERATION..............................................................

199 6.4 ACTION TO BE TAKEN IF A SAFETY LIMIT IS E1CEEDED.............................................

199 6.5 PLANT OPERATING PROCEDURES...................................................................

200 6.6 PLANT OPERATING RECORDS......................................................................

207 6.7 PLANT REPORTING REQUIREMENTS...............................................,.................

208 6.8 FIRE PROTECTION INSPECTION...................................................................

216 6.9 ENVIRONMENTAL QUALIFICATION..................................................................

216 6.10' INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT.....................................................

217 6.11 IODINE MONITORING............................................................................

217 6.12 PROCESS CONTROL PROGRAM (PCP)................................................................

217 6.13 OFF-SITE DOSE CALCULATION MANUAL (ODCM)......................................................

218 6.14 MAJOR CHANGES TO RADIOACTIVE LIQUID, CASEOUS, AND SOLID WASTE TREATMENT SYSTEMS...........................................................

219 8503200269 850304 g

Amendment No. 61 re u foy m ggq6fq23

TABLE OF CONTENTS (Cont'd)

I LIMITING SAFETY SAFETY LIMITS Page No.

SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY.............................................

S 2.1 1.2 REACTOR COOLANT SYSTEM.....................'.........................

16 2.2 LIMITING CONDITIONS OF OPERATION Page No.

SURVEILLANCE 3.1 REACTOR PROTECTION SYSTEM...........................................

18 4.1 BASES 27 3.2 PROTECTIVE INSTRUMENT SYSTEMS.......................................

32 4.2 A.

Emergency Core Cooling System...................................

32 A

B.

Primary Containment Isolation...................................

33 B

C.

Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation.................................

33 C

D.

Air Ejector Off-Gas System Isolation............................

33 D

E.

Control Rod Block Actuation.....................................

33 E

F.

Mechanical Vacuum Pump Isolation................................

34 F

G.

Post-Accident Instrumentation...................................

34 G

H.

Drywell to Torus P Instrumentation.............................

34 H

I.

Recirculation Pump Trip Instrumentation.........................

34a.................

I l

BASES 62 3.3 CONTROt ROD SYSTEM..................................................

68 4.3 A.

Reactivity Limitations..........................................

68 A

B.

Control Rods....................................................

69 B

C.

Scram Insertion Times...........................................

72 C

D.

Control Rod Accumulators........................................

73 D

E.

Reactivity Anomalies............................................

74 E

BASES 75 l

3.4 REACTOR STANDBY LIQUID CONTROL SYSTEM...............................

79 4.4 A.

Normal Operation................................................

79 A

B.

Operation with Inoperable Components............................

80 B

C.

Liquid Poison Tank - Boron Concentration........................

80 C

BASES 83 Amendment No. 67 1

TABLE OF CONTENTS (Cont'd)

LIMITING CONDITIONS OF OPERATION Page No.

SURVEILLANCE 3.5 CORE AND CONTAINENT COOLING SYSTEMS................................

85 4.5 A.

Core Spray and Low Pressure Coolant Inj ection...................

85 A

B.

Containment Spray Cooling Capability............................

87 B

C.

Residual Heat Removal (RNR) Service Water System................

88 C

D.

Station Service Water and Alternate Cooling Tower Systems.......

89 D-E.

High Pressure Coolant Inj ection (NPCI) System...................

91 E

F.

Automatic Depres surization System................................

92 F

G.

Reactor Core Isolation Cooling System (PCIC).....................

93 G

H.

Minimum Core and Containment Cooling System Availability.........

94 H

I.

Maintenance of Filled Discharge Pipe.............................

95 I

l BASES 99 3.6 REACTOR COOLANT SYSTEM...............................................

105 4.6 A.

Pres sure and Temperature Limitations.............................

105 A

B.

Coolant Chemistry................................................

106 B

C.

Coolant Leakage..................................................

108 C

D.

Safety and Relief Va1ves.........................................

108 D

E.

Structural Integrity.............................................

109 E

F.

Jet Pumps........................................................

109 F

G.

Single Loop Operation............................................

110 H.

Recirculation System.............................................

110 I.

Shock Suppressors................................................

110s.................

I J.

Thermal Hydraulic Stability......................................

110b.................

J I

BASES 117 3.7 STATION CONTAINMENT SYSTEMS..........................................

126 4.7 A.

Primary Containment..............................................

126 A

B.

Standby Cas Treatment............................................

130 B

C.

Secondary Containment System.....................................

131 C

D.

Primary Containment Isolation Va1ves.............................

132 D

BASES 138 3.8 RADIOACTIVE EFFLUENTS................................................

147 4.8 A.

Liquid Effluents:

Concentration.................................

147 A

B.

Liquid Effluents:

Dose..........................................

148 B

C.

Liquid Radweste Treatment........................................

149 C

D.

Liquid Holdup Tanks..............................................

149 D

E.

Gaseous Effluents:

Dose Rate....................................

150 E

Amendment No. A3, 57, 83 mfuW

TABLE OF CONTENTS (Cont'd)

]

LIMITING CONDITIONS OF OPERATION Page No.

SURVEILLANCE 1

s F.

Gaseous Effluents: Dose From Boble Cases........................

151 F

G.

Gaseous Effluents: Dose From Iodine-131. Iodine-133. Tritium,:

and Radioactive Materials in Particulate Fora....................

-152-G H.

Gaseous Radweste Treatment.......................................

152 H

I..

Ventilation Exhaust Treatment....................................

153 I

J.

Explosive Gas Mixture............................................

153-J K.

Steam Jet Air Ej ec to r ( SJAE).....................................

153 K

L.

Primary Containment..............................................

154 L

M.

Total Dose.......................................................

155 M

N.

Solid Radioactive Waste..........................................

156 N

i BASES 160d i

3.9 RADIOACTIVE EFFLUENT MONITORING SYSTEMS..............................

161 4.9.

A.

Liquid Effluent Monitoring System................................

161 A

B.

Gaseous Effluent Instrumentation.................................

161a.................

B C.

Radiological Environment Monitoring Program......................

161a.................

C

{

D.

Land Use Census..................................................

162 D

E.

Intercomparison Program..........................................

163 E

BASES 172j l

3.10 AUKILIARY' ELECTRICAL POWER SYSTEMS...................................

173 4.10 1

I A.

Normal Operation.................................................

173 A

}

B.

Operation with Inoperable Components.............................

176 B

C.

Diesel Fuel......................................................

177 C

4 l

l BASES 178 l

3.11 REACTOR FUEL ASSEMBLIES..............................................

180s.................

4.11 A.

Average Planar LHGR..............................................

180s.................

A B.

LHCR.............................................................

180b.................

B l

C.

MCPR.............................................................

180c.................

C BASES 180d j

3.12 REFUELING AND SPENT FUEL HANDLING....................................

181 4.12 I.

A.

Refueling Interlocks.............................................

181 A

B.

Core Monitoring..................................................

182 B

C.

Fuel Storage Pool Water Level....................................

183 C

1 1

i 1

j Amendment No. A3, 67, 83

)

-%1-

TABLE OF CONTENTS (Cont'd)

LIMITING CONDITIONS OF OPERATION Page No.

SURVEILLANCE D.

Control Rod and Control Rod Drive Maintenance....................

184 D

E.

Extended Core Maintenance........................................

184 E

F.

Fuel Movement....................................................

185 F-r G.

Crane Operability................................................

185 G

H.

Spent Fuel Pool Water Temperature................................

185a.................

H BASES 186-3.13 FIRE PROTECTION SYSTEM...............................................

187b.................

4.13 BASES 187m 5.0 DESIGN FEATURES......................................................

188 I

6.0 ADMINISTRATIVE CONTROLS..............................................

190 i

4 i

t v

l 1

4 4

l l

't 1

1 1

Amendment No. A3, 87, 83

+

VYWpS 1.0 DEFINITIONS F.

Instrument check - An instrument check is qualitative determination,of acceptable i

The succeeding frequently used terms are operability by observation of instrument explicitly defined so that a uniform.

behavior during operation. This interpretation of the specifications may be determination shall include, where possible, achieved.

comparison of the instrument with other independent instruments measuring the same A.

Reportable Event - A reportable event shall variable.

be any of these conditions specified in j

Section 50.73 to 10CFR Part 50.

B.

Alteration of the Reactor Core - The act of moving any component in the region above the core support plate, below the upper grid and within the shroud. Normal movement of the control rods, or the neutron detectors is not defined as a core alteration.

l C.

Hot Standby - Hot standby means operation I

with the reactor critical and the main steam j

line isolation valves closed.

I D.

Immediate - Immediate means that the required action will be initiated as soon as j

practicable considering the safe operation of J

the unit and the importance of the required action.

1 E.

In'stnament Calibration - An instrument l

calibration means the adjustment of an instrument signal output so that it I

corresponds, within acceptable range and j

accuracy, to a known valve (s) of the parameter which the instrument monitors, j

Calibration shall encompass the entire j

instrument including actuation, alarm, or trip. Response time as specified is not part i

of the routine instrument calibration but will be checked once per operating cycle.

l Amendment No. 28, 1

_.. ~..

VYNFS 3.2 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS C.

Liquid Radweste Treatment C.

Liquid Radweste Treatment 1.

The liquid radweste treatment system 1.

See Specification 4.8.B.1.

shall be used in its designed modes of operation to reduce the radioactive materials in the liquid waste prior to its discharge when the estimated doses due to the liquid effluent from the site, when averaged with all other liquid release over the last month, would exceed 0.06 meen to the total body, or 0.2 arem to any organ.

D.

Liquid Holdup Tanks D.

Liquid Holdup Tanks 1.

The quantity of radioactive material 1.

The quantity of radioactive material contained in any outside tank

• shall be contained in each of the liquid holdup Limited to less than or equal to 10 tanks

• shall be determined to be within curies, excluding tritium and dissolved the limits of Specification 3.8.D.1 by or entrained noble gases, analyzing a representative sample of the tank's content at least once per week l

when radioactive materials are being added to the tank.

• NOTE: Tanks included in this Specification are only those outdoor tanks.that are not surrounded by liners, dikes, or walls capable of holding the tank's contents, or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

Amendment No. 83 149

VYNPS

-s Bases:

3.11C Minimum Critical Power Ratio (MCPR)

_Operatina Limit MCPR 1.

The MCPR Operating Limit is a cycle-dependent parameter which can be determined for a number of different combinations of operating modes, initial conditions, and cycle exposures in order to provide reasonable assurance against exceeding the Fuel Cladding Integrity Safety Limit (FCISL) for potential abnomal occurrences. The MCPR operating limits are justified by the analyses, the results of which are presented in the current cycle's Core Performance Analysis Report.

Amendment No. A1, le, 83 180h

VYNPS 2.

Annual Report

~

j 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 meen/yr and their associated man-rem exposure according to work and job functions,1/ e.g.,

reactor operations and curveillance, inservice inspection, routine maintenance, special maintenance l

(describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD or film badge measurement. Small j

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, i

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 l

describes the operation of the facility.

9 Reportable Occurrences 1

Reportable occurrences, including corrective actions and measures to prevent recurrence, shall be reported to the NRC in accordance with the requirements of Section 50.73 to 10CFR Part 50.

Events I

involving systems or components described in Sections 3/4.8.B. 3/4.8.C, 3/4.8.F. 3/4.8.G, 3/4.8.H, i

3/4.8.I. 3/4.8.M. 3/4.9.C, 3/4.9.D. 3/4.9.E Table 3.9.1-note 5. Table 3.9.2-note 7, and 3/4.13 do not i

require reporting under the provision of this sectioa. Such events will be reported as required in Section 6.7.C.2 or 6.7.C.3 as indicated below. The reporting provisions of this section are not applicable to Sections 3/4.8.A. 3/4.8.D 3/4.8.E, 3/4.8.N, 3/4.9.A. and 3/4.9.B.

1/ This tabulation supplements the requirements of 20.407 of 10CFR Part 20.

Amendment No. A2, 83 209

VYNPS C.

Unique Reportina Requirements 1.

Semiannual Effluent Release Report a.

Within 60 days after January 1 and July 1 of each year, a report si.all be submitted covering the radioactive content of effluents released to unrestricted areas during the previous six months of operation.

b.

The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, Revision 1. June 1974, " Measuring, Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Caseous Ef fluents from Light-Water-Cooled Nuclear Power Plants", with data summarized on a i

quarterly basis following the format of Appendix B thereof. For solid wastes the format for Table 3 in Appendix B of Regulatory Guide 1.21 shall be supplemented with three additional categories: class of solid wastes (as defined by 10CFR Part 61), type of container (e.g., LSA, Type A, Type B, Large Quantity), and solidification agent or absorbent, if any.

1 In addition, the radioactive effluent release report to be submitted 60 days after January I

1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This anntal summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.* This same report (or a supplement to it to be submitted within 180 days of January 1 each year) shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit during the previous calendar year.

(The semiannual effluent release report submitted j

within 60 days of July 1 each year need not contain any dose estimates from the previous 6 months' effluent releases.) The effluent reported submitted after January 1 each year shall also include an assessment of the radiation doses from radioactive effluents to 4

eIn lieu of submission with the first half year radioactive effluent release report, the licensee has the options of retcining this summary of required meteorological data in a file that shall be provided to the NRC upon request.

Amendment No. 83 210 i

l l

i

VYMPS h r(s) of the public due to any allowed recreational activities inside the site boundary

)

during the previous calender year. All assumptions used in making these assessments (e.g.,

specific activity, exposure time and location) shall be included in these reports. For any batch or discrete gas volume releases, the meteorological conditions concurrent with the j

time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. For l

radioactive materials released in continuous effluent streams, quarterly average I

meteorological conditions concurrent with the quarterly release period shall be used for l

determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the off-site Dose Calculation Manual (ODCM).

l With the limits of Specification 3.8.M.1 being exceeded during the calendar year, the I

radioactive effluent release report to be submitted 60 days after January 1 of each year l

shall also include an assessment of radiation doses to the likely most exposed real member (s) of the public from reactor releases (including doses from primary effluent l

pathways and direct radiation) for the previous calendar year to show conformance with I

40CFR190, Environmental Radiation Protection Standards for Nuclear Power Operation.

The radioactive effluent release reports shall include a list and description of unplanned releases from the site to site boundary of radioactive materials in gaseous and liquid

}

effluents made during the reporting period.

I J

With the quantity of radioactive material in any outside tank exceeding the limit of Specification 3.8.D.1, describe the events leading to this condition in the next j

Radioactive Effluent Release Report.

}

l If inoperable radioactive liquid effluent monitoring instrumentation is not returned to operable status prior to the next release pursuant to Note 4 of Table 3.9.1, explain in the next Radioactive Effluent Release Report the reason (s) for delay in correcting the inoperability.

[

If inoperable gaseous effluent monitoring instrumentation is not returned to operable status within 30 days pursuant to Note 5 of Table 3.9.2, explain in the next Radioactive Effluent Release Report the reason (s) for delay in correcting the inoperability.

211 Amendment No. 83 lI

V With milk samples no longer available from one or more of the sample locations required by~

Table 3.9.3, identify the cause(s) of the sangle(s) no longer being available, identify the 4

j new location (s) for obtaining available replacement samples, and include revised ODCM l

.j figure (s) and table (s) reflecting the new location (s) in the next Radioactive Effluent Release Report.

'f With a land use census identifying one or more locations which yield at least a 20 percent 4

i greater dose or dose comunitment than the values currently being calculated in Specification 4.8.C.1, identify the new location (s) in the next Radioactive Effluent Release Report.

l l

Changes made during the reporting period to the Process Control Program (PCP) and to the i

Off-Site Dose Calculation Manual (ODCM), shall be identified in the next Radioactive

)

Effluent Release Report.

1 I

2.

Special Reports

}

Special reports shall be submitted to the Director of the Office of Inspection and Enforcement i

Regional Office within the time period specified for each report.

a.

Liquid Effluents, Specifications 3.8.B and 3.8.C.

4 l

With the calculated dose from the release of radioactive materials in liquid effluents j

exceeding any of the limits of Specification 3.8.B.1, prepare and submit to the Connaission l

within 30 days a.special report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions taken to assure that subsequent releases will be in compliance with the limits of Specification 3.8.B.1.

l With liquid radweste being discharged without processing throdsh appropriate treatment systead and estimated doses in excess of Specification 3.8.C.1, prepare and submit to the 4

l Comunission within 30 days a special report which includes the following information:

(1) explanation of why liquid radwaste was being discharged without treatment, 4

J identificatioa of any inoperable equipment or subsystems, and the reasons for the inoperebilitj;

[

i l

Amendment No. 83 212 i

i i

_,____..m, r.__.r_____..____.,..,._,y,

..,,.,,...__,_.,,,,_,....._,,.m.

,m...

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

VYWPs l

l (2) action (s) taken to restore the inoperable equipment to operable status; and (3)

Summary description of action (s) taken to prevent a recurrence.

f b.

Gaseous Effluents. Specifications 3.8.F. 3.8.G. 3.8.H. and 3.8.I.

i With the calculated air dose from radioactive noble gases in gaseous effluents exceeding i

any of the limits of Specification 3.8.F.1, prepare and submit to the Comunission within 30 i

days a special report which identifies the cause(s) for exceeding the limit (s) and the corrective action (s) taken to assure that subsequent releases will be in compliance with 1

the limits of Specification 3.8.F.1.

With the calculated dose from the release of Iodine-131. Iodine-133, tritium, and/or radionuclides in particulate forin exceeding any of the limits of Specification 3.8.G.1, prepare and submit to the Commission within 30 days a I

special report which identifies the cause(s) for exceeding the limit (s) and the corrective action (s) taken to assure that subsequent releases will be in compliance with the limits of 1,

specification 3.8.G.I.

)

With gaseous radweste being discharged without processing through appropriate treatment j

systems as defined in Specification 3.8.H.1 for more than seven (7) consecutive days, or in excess of the limits of Specification 3.8.I.1, prepare and submit to the Comunission within 30 days a special report which includes the following information:

l I

(1) explanation of why gaseous radweste was being discharged without treatment i

(Specification 3.8.H.1), or with resultant doses in excess of Specification 3.8.I.1, j

identification of any inoperable equipment or subsystems, and the reasons for the j

inoperability; i

l (2) action (s) taken to restore the inoperable equipment to operable status; and l

(3) summary description of action (s) taken to prevent a recurrence, i

i l

l Amendment No. 83 213 i

4

VYNPS c.

Total Dose, Specification 3.8.M.

1

?

l 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 1

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 1

report, required by 10CFR Part 20.405c, shall include an analysis that estimates the l

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 I

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 j

granted until staff action on the request is complete.

d.

Radiological Environmental Monitoring, Specification 3.9.C.

1 l

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 I

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 l

was not the result of plant effluents, however, in such an event, the condition shall be i

reported and described in the annual Radiological Environmental Surveillance Report.

e.

Land Use Census, Specification 3.9.D.

With a land use census not being conducted as required by Specification 3.9.D, prepare and l

submit to the Commission within 30 days a special report Which identifies the reasons why the survey was not conducted, and what steps are being taken to correct the situation.

1 I

i 4

N I

j Amendment No. 83 214 l

_._ _ _ _ _ _ _ _ j

w 0

f.

Vital Fire Protection System Specification 3.13 Where required by section 3.13,.special reports shall be submitted to the Commission following the discovery of certain inoperable sensors, instrument.s, components, or systems in the vital fire protection system.

Routine surveillance testing or design modification of sensors, instruments, Note:

components, or systems which lead to operation of sensors, instruments,' components, or systems in a degraded mode do not require special reporting except where tests, themselves reveal a degraded mode.

3.

Environmental Radiolozical tionitorinz Radiological Environmental Surveillance Reports covering the operation of the unit during previous calendar year shall be submitted prior to May 1 of each year.

The annual Radiological Environmental Surveillance Report shall include summaries, interpretations, and an analysis of trends of the results of the radiological onvironmental surveillance activities for the report period, including a comparison with operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

The annual Radiological Environmental Surveillance Report shall include sununarized and tabulated

}

results of all radiological environmental samples taken during the report period pursuant to the l

table and figures in the ODCM. In the event that some results are not available for inclusion l

with the report, the report shall be submitted noting and explaining the reasons for the missing j

results. The missing data shall be submitted as soon as possible in a supplementary report.

?

With the level of radioactivity 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, the condition shall be described in the next annual Radiological Environmental Surveillance Report only if the I

measured level of radioactivity was not the result of plant effluents. With the radiological l

environmental monitoring program not being conducted as specified in Table 3.9.3, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence shall be included in the next annual Radiological Environmental Surveillance Report.

kmendment No. 83 215

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- ~ ' "

VYNPS The annual Radiolcgical Environmentcl Suc*vaillanto R; port chall also include the results of the land use census required by Specification 3.9.D.

A summary description of the radiological i

environmental monitoring program including a map of all sampling locations keyed to a table

}

giving distances and directions from the reactor shall be in the reports.

If new environmental sampling locations are identified in accordance with Specification 3.9.D, the new locations shall be. identified in the next annual Radiological Environmental Surveillance Report.

1 The reports shall also include a discussion of all analyses in which the LLD required by Table 4.9.3 was not schievable.

~

The results of licensee participation in the intercomparison program required by Specification 3.9.E shall'be included in the reports. With analyses not being performed as i

required by Specification 3.9.E. the corrective actions taken to prevent a recurrence shall be i

report to the Counission in the next annual Radiological Environmental Surveillance Report.

4 6.8 FIRE PROTECTION INSPECTION An independent fire protection and loss prevention inspection and audit shall be performed annually A.

utilizing either qualified off-site licensee personnel or an outside fire protection firm.

4 B.

An inspection and audit by an outside fire consultant shall be performed at intervals no greater than 3 years.

6.9 ENVIRONMENTAL QUALIFICATION i

A.

By no later than June 30, 1982, all safety-related electrical equipnent in the facility shall be qualified in accordance with the provisions of Division of Operating Reactors, " Guidelines for j

Evaluating Environmental Qualification of Class 1E Electrical Equipmcat in Operating Reactors" (DOR Guidelines); or NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related j

Electrical Equipnent", December 1979. Copies of these documents are attached to Order for

)

Modification of License DPR-28, dated October 24, 1980.

)

B.

By no later than December 1, 1980, complete and auditable records must be available and maintained at j

a central location which describe the environmental qualification method used for all safety-related l

electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines i

or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is l

replaced, further tested, or otherwise qualified.

l Amendment No. 83 216 i

YYEPS 6.10 teamssITY OF ii ansa. OUTSIDE COETAIMEENT A program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels will be implemented. This program shelt be3.'*de the following:

Provisions establishing preventive maintenance and p.eriodic visual inspection requirements.

A.

B.

System leakage inspections, to the extent permitted by system design and radiological conditions, for each system at a frequency not to exceed refueling cycle intervals. The systems subject to this testing are:

(1) Residual Heat Removal, (2) Core Spray, (3) Reactor' Water Cleanup. (4) HPCI, (5) RCIC, and (6) Sampling Systems.

6.11 IODINE MONITORING A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas

• under accident conditions will be implemented. This program shall include the following:

A.

Training of personnel.

B.

Procedures for monitoring.

C.

Provisions for maintenance of sampling and analysis equipment.

6.12 PROCESS CONTROL PROGRAM (PCP)

A process control program shall contain the sampling, analysis, tests, and determinations by which wet radioactive waste from liquid systems is assured to be converted to a. form suitable for off-site disposal..'

A.

Licensee initiated changes to the PCP:

• Areas requiring personnel access for establishing hot shutdown condition.

Amendment No. A3, 83, 83 217

j YYWPS 4

i 1.

Shall be submitted to the Consnission in the semiannual Effluent Release Report for the period in which the change (s) was made. This submittal'shall contain:

l a.

Sufficiently detailed information to support the rationale for the change without benefit l

of additional or supplemental information.

l I

b.

A determination that the change did not reduce the'overall conformance of the dowatered spent resins / filter media waste product to existing criteria for solid waste shipmen,ts and disposal.

r

}

Documentation of the' fact that the change has been reviewed by PORC and approved by the c.

Manager of Operations (M00).

j 2.

Shall become effective upon review by PORC and approval by the Manager of Operations (M00).

1 1

j 6.13 0FF-SITE DOSE CALCULATION MANUAL (ODCM) i 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, 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 Comnission in the semiannual Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

(

I Sufficiently detailed information to support the rationale for the change without benefit !

a.

of additional or supplemental information. Information submitted should consist of a f

package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s).

1 b.

A determination that the change will not reduce the accuracy or reliability of <*ane calculations or setpoint determinations.

Amendment No. 83, 83 218

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VYNPS Documentation of the fact that the change hAs been reviewed by PORC and approved by the c.

Manager of Operations (M00).

2 2.

Shall become effective upon review by PORC and approved by the Manager of Operations (M00).

i 6.14 MAJOR CHANGES TO RADIOACTIVE LIOUID. r2A*m39. aun w Tn WASTE T***TMENT SYSTEMS

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

I j

A.

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

4 j

1.

A sunnary of the evaluation that led to the determination that the change could be made in

,accordance with 10CFR Part 50.59; i

2.

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

3.

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

)

4.

An evaluation of the change, which shom: the predicted releases of radioactive materials in i

liquid and gaseous effluents and/or quantity of solid waste that differ from those previously j

predicted in the license application and amendments thereto; I

j 5.

An evaluation of the change, which shows the expected maximum evposures to member (s) of the j

public at the site boundary and to the general population that.!.ffer from those previously i

estimated in the license application and amendments thereto; f

f j

6.

A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; l

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

i

\\

Amendment No. 83 219 I

i

. ~.

VYEPS 7.

An 00timato cf the exposure to plant opelsting personnel CD a result of the change; and i

8.

Documentation of the fact that the change was reviewed and found acceptable by PORC.

i B.

Shall become effective upon review and acceptance by PORC and approval by the Plant Manager.

1 i

.i a

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i i

I 1

i

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Amendment No. 83 220 I

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