ML20094R271

Amend 174 to License DPR-46,revising TS to Include Wording Consistent W/Revised 10CFR20

ML20094R271
Person / Time
Site:	Cooper
Issue date:	11/28/1995
From:	Hall J NRC (Affiliation Not Assigned)
To:
Shared Package
ML20094R273	List: ML20094R271 ML20094R302
References
NUDOCS 9512040340
Download: ML20094R271 (16)

Text

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UNITED STATES i

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2066H001 o

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NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 174 License No. DPR-46 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Nebraska Public Power District (the licensee) dated June 14, 1993, as supplemented by letter dated 1

April 12, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(1) 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 i

conducted in compliance with the Commission's regulations; D.

The issuance of this license 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.

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1 9512040340 951128 PDR ADOCK 05000298 P

PDR

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-46 is hereby amended to read as follows:

2.

Technical Specifications The Technical Specifications contained in Appendix A, as revised 3

through Amendment No.174, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Nnd

/

1 Ja g R. Hall, Senior Project Manager P.oject Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: November 28, 1995 5

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4 ATTACHMENT TO LICENSE AMENDMENT NO.174 FACILITY OPERATING LICENS[ NO. DPR-46 DOCKET NO. 50-298 Replace the following pages of the Appendix A Technical Specifications with I

the enclosed pages. The revised pages are identified by Am=dment number and i

contain vertical lines indicating the area of change.

REMOVE PAGES INSERT PAGE5 11 11 iv iv Sa 5a 185 1

186 187, 188, 189, 190, 191, 192 185, 186, 187, 188, 189, 190, 191, 192 1

216x 216x 216z 216z 216a8 216a8 216a21 216a21 216a22 216a22 216a26 216aN 1

226 226 229 229 l

231 231 a

4 4

e TABLE OF CONTENTS (cont'd)

Pare No.

SURVEILIANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 114 - 131 A.

Core Spray and LPCI Systems A

114 B.

RHR Service Water System B

116 C.

HPCI System C

117 D.

RCIC System D

118 E.

Automatic Depressurization System E

119 F.

Minimum Low Pressure Cooling System Diesel Generator Availability F.

120 C.

Maintenance of Filled Discharge Pipe C

122 3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 A.

Thermal and Pressurization Limitations A

132 B.

Coolant Chemistry B

133a C.

Coolant Leakage C

135 D.

Safety and Relief Valves D

136 E.

Jet Pumps E

137 F.

Recirculation Pump Flow Mismatch F

137 C.

Inservice Inspection G

137 H.

Shock Suppressors (Snubbers)

H 137a 3.7 CONTAINMENT SYSTEMS 4.7 159 - 192 A.

Primary Containment A

159 B.

Standby Cas Treatment System B

165 C.

Secondary Containment C

165a D.

Primary Containment Isolation Valves D

166 3.8 DELETED 185 - 186 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 193 - 202 A.

Auxiliary Electrical Equipment A

193 B.

Operation with Inoperable Equipment B

195 3.10 CORE ALTERATIONS 4.10 203 - 209 A.

Refueling Interlocks A

203 B.

Core Monitoring B

205 C.

Spent Fuel Pool Water Level C

205 D.

Time Limitation D

206 E.

Spent Fuel Cask Handling E

206 3.11 FUEL RODS 4.11 210 - 214e 4

A.

Average Planar Linear Heat Generation Rate (APLHGR)

A 210 B.

Linear Heat Generation Rate (LHGR)

B 210 C.

Minimum Critical Power Ratio (MCPR)

C 212 Amendment No. Oj,07,100,152,100,ITA,

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TABLE OF CONTENTS (Cont'd.)

Pa ge Eo.

SURVEILIANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 6.2 Review and Audit 220 6.2.1.A Station Operations Review Committee (SORC) 220 A.1 Membership 220 J

A.2 Heeting Frequency 220 i

A.3 Quorum 220 A.4 Responsibilities 220

)

A.5 Authority 221 I

i A.6 Records 221 A.7 Procedures 221 6.2.1.5 NPPD Safety Review and Auoit Board (SRAB)

.'22 B.1 Membership 2D j

B.2 Meeting Frequency 222 i

B.3 Quorum 222

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B.4 Review 222 i

B.5 Authority 223 B.6 Records 223 B.7 Audits 223 6.3 Procedures and Programs 225 6.3.1 Introduction 225 6.3.2 Procedures 225 6.3.3 Maintenance and Test Procedures 225 6.3.4 Radiation Control Procedures 225

.A High Radiation Areas 226 6.3.5 Temporary Changes 227 6.3.6 Exercise of Procedures 227 6.3.7 Programs 227

.A Systems Integrity Monitoring Program 227

.B Iodine Monitoring Program 227

.C Post-Accident Sampling System (PASS) 227 6.4 Record Retention 228 6.4.1 5 year ret aon 228 6.4.2 Life reter ion 228 6.5 Station Reporting Re.quirements 230 6.5.1 Routine Reports 2 30

.A Introduction 230

.B Startup Report 230

.C Annual Reports 230

.D Monthly Operating Report 231

.E Annual Radiological Environmental Report 231

.F Semiannual Radioactive Material Release Report 2 32

.C Core Operating Limits Report 2 32 Amendment No.195d42,174

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W.A Solidification - SOLIDIFICATION shall be the conversion of radioactive wastes j

from liquid systems to a solid which is as uniformally distributed as i

reasonably achievable with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

X.

Spiral Raioad - Pertains to the spiral reloading of the core with fuel, at I

j least. 50% of which has previously accumulated a minimum exposure of 1000 NUD/T.

i Y.

Eurveillance Freauenev - Surveillance requirements shall be applicable during the operational conditions associated with individual 140's unless otherwise stated in an individual Surveillance Requirement.

l Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of j

the specified surveillance interval.

i Performance of a Surveillance Requirement within the specified time interval shall constitute compliance with operability requirements for an 140 unless i

otherwise required by the specification.

i The Surveillance Frequency establishes the limit for which the specified time interval for Surveillance Requirements may be extended.

It permits an-i 4

1 allowable extension of the normal surveillance interval to facilitate surveillance schedule and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g.,

transient conditions or other ongoing surveillance or maintenance activities. It also j

provides flexibility to accommodate the length of a fuel cycle for surveillance that are performed at each refueling outage and are specified i

a i

with an 18 month surveillance interval.

It is not intended that this l

provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for survefilances that are not performed l.

during refueling outages.

The limitation of this definition is based on engineering judgement and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ansured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance i

j interval.

2.

Surveillance Interval The surveillance interval is the calendar time between surveillance tests, checks, calibrations and examinations to be performed upon an instrument or component when it is required to be operable.

i These tests may be waived when the instrument, component or system is not I

j required to be operable, but the instrument, component or system shall be l

tested prior to being declared operable or as practicable following its j

j, return to service.

Z.A Venting - Venting is the controlled process of discharging air or gas from a confinement to establish temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is i

i not provided or required during venting.

Vent, used in system names, does j

not imply a venting process.

2.B Offsite - Offsite means outside of the exclusion area as defined in 10CFR l

Part 100.3.

TheexclusionareaboundaryaroundCooperNuclearStationisl defined in Figure 1.1 and may also be referred to as the Site Boundary.

l 2.C Member of the Public - A Member of the Public is a person in a controlled or unrestricted area who does not receive an occupational dose.

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Amendment No. M4,174

-Sa-i-

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" INTENTIONAL 1Y 12FT BLANK"

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- Amendemnt No. T/4"^ 15 t,

-185,.186, 187, 188, 189, 190, 191, 192-i u

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2 LIMITING CONDITION FOR OPERATION SURVEILIANCE REOUIRENF.NTS 3.21 (Cont'd) 4.21 (Cont'd)

B.

Liquid Effluents B.

Liould Effluents Applicability: At all times.

1.

Concentration Specification:

a.

Radioactive liquid wastes shall be 1

sampled and analyzed according to I

j 1.

Concentration Table 4.21.B.1.

j a.

He concentration of radioac-b.

The analytical results shall be used i

tive material in water with methods in the ODAM to verify l

OFFSITE (Figure 1.1) due to that the average concentration be-l 1

radioactive liquid affluent yond the SITE BOUNDARY does notl

)

i shall not exceed the concen-exceed Specification 3.21.B.1.a.

tration eyecified in 10 CPR when Sr-89, Sr-90 and Fe-55 concen-l Fart 20.1302 for trations are averaged over no more 3

l radionuclides other than than 3 months and other radionuclide l

dissolved or entrained noble concentrations are averaged over no Bases.

For dissolved or more than 31 days.

l entrained noble gases, the i

concentration shall not exceed 2 x 10 pCi/ml total activity.

i b.

With the concentration of l

radioactive material released l

OFFSITE exceeding the limit, attend to the cause without i

j delay and restore the concen-j tration within the above lin-l it.

l c.

He provisions of Specifica-l tion 6.5.2 do not apply.

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i Amendment No. 49,174

-216x-i

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i NOTES FOR TABLE 4.21.B.1 (1) The LIE is the smallest concentration of the radioactive material in a sample that will be detected with 951 probability (51 probability of falsely concluding that a j

blank observation represents a "real" signal).

l For a particular measurement system (which may include radiochemical separation):

l 4.66 s I

b

~

E

• V

• 2.22

• Y

• exp (-A&t) l Where:

j LIE is the "a priori" lower limit of detection as defined above (as picoeurie per unit mass or volume),

2 s is the standard deviation of the background counting rate or of the counting u

rIts of a blank naple as appropriate (as counts per minute),

E is the counting efficiency (as counts per transformation),

V is the sample size (in units of mass or volume),

2.22 is the number of transformations per minute per picoeurie, Y is the fr actional radiochemical yield (when applicable),

4 1 is t'a radioactive decay constant for the particular radionuclide, and at is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

a The value of s used in the calculation of the LLD for a detection system shall b

l be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance.

In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the

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samples. Typical values of E. V, Y, and At shall be used in the calculation.

(2) For certain radionuclides with low gamma yield or low energies, or for certain i

radionuclide mixtures, it may not be possible to measure radionuclides in concentra-i tions near the LLD. Under these circumstances, the LLD may be i eressed inversely j

proportionally to the magnitude of the gamma yield (i.e., 5 x 10"p/I, where I is the photon abundance expressed as a decimal fraction), but in no case shall the LID, as i

calculated in this manner for a specific radionuclide, be greater than 10% of the values specified in 10 CFR 20, Appendix B, Table 2, Column 2.

(3) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed

_ results in a specimen which is representative of the liquids released.

(4) To be representative of the quantities and concentrations of radioactive materials in liquid affluents, daily grab samples shall be collected in proportion to the rate of flow of the effluent stream.

Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the affluent release.

Amendment No. 49,174

-216z-

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NOTES FOR TABLE 4.21.C.1 4

1.

(1) The LIA is the smallest concentration of radioactive material in a sample that will be detected with 951 probability (51 probability of falsely concluding that a blank observation represents a "real" signal).

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J For a particular measurement system (which may include radiochemical separation):

l

.4.66 sb

~

)

E. V. 2.22. Y. exp (-AAt) l Where-i LLD is the "a priori" lower limit of detection as defined above (as picocuria i

l per unit mass or volume),

1 s is the standard deviation of the background counting rate or of the counting u

rite of a blank sample as appropriate (as counts per minute),

)

i E is the counting efficiency (as counts per transformation),

[

V is the sample size (in units of mass or volume),

j 2.22 is the number of transformations per minute per picoeurie, a

j Y is the fractional radiochemical yield (when applicable),

1 is the radioactive decay constant for the particular radionuclide, and 3

At is the elapsed time between midpoint of sample collection and time of l

counting (for plant effluents, not environmental samples).

t The value of s used in the calculation of the LLD for a detection ' system shall l

l be based on thI actual observed variance of the background counting rate or of

}

the counting rate of the blank samples (as appropriate) rather than on an j

unverified theoretically predicted variance.

In calculating the LLD for a 4

l radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the I

samples. Typical values of E, V, Y, and At shall be used in the calculation.

i (2) For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentra-tions near the LLD. Under these circumstances, the LLD may be j

proportionaltothemagnitudeofthegammayield(i.e.,1x10'j/I,whereIisthencreased inversely i

J photon abundance expressed as a decimal fraction), but in no case shall the LLD, as i

calculated in this manner for a specific radionuclide, be greater than 10% of the values specified in 10 CFR 20, Appendix B. Table 2, Column 1.

1 I

(3) Analyses shall also be performed following an increase as indicated by the gaseous release monitor of greater than 501 in the steady state release, after factoring out i

increases due to power changes or other operational occurrences, which could alter 1

the mixture of radionuclides.

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Amendment No. 49,174

-216a8-i i

i-i 3.21 & 4.21 AMER 3.21.A & 4.21.A INSTRUMENTATION 3.21.A.1 & 4.21.A.1 Liouid Effluent Monitorinn 4

The radioactive liquid affluent instrumentation is provided to monitor and control, as applicable, the release of radioactive material in liquid effluents. The OPERABILITY and use of these instruments implements the requirements of 10 CFR Part 50, Appendix A, General Design Criteria 60, 63, and 64.

The alarm and/or trip setpoints for these instruments are j

calculated in the manner described in the ODAM to assure that the alarm and/or trip will occur before the limit specified in 10 CFR Part 20.1302 is exceeded. Control of the normal l j

liquid discharge pathway is assured by station procedures governing locked discharge valves and valve line-up verification.

l 3.21.A.2 & 4.21.A.2 Gaseous Effluent Monitorinn The radioactive gaseous affluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The location of this instrumentation is indicated 4

by a Figure in the ODAM, a simplified flow diagram showing gaseous effluent treatment and i

monitoring equipment. The alarm / trip setpoints for these instruments shall be calculated in accordance with methods in the ODAM, which have been reviewed by NRC, to ensure that the alarm will occur prior to exceeding the limits of 10 CFR Part 20.

The process monitoring instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the augmented offgas treatment system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, j

and 64 of Appendix A to 10 CFR Part 50.

In the event no flow rate measurement device is operable on a gaseous stream, alternative 24-hour estimates are adequate since the system design is constant flow and loss of flow is

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clarmed in the control room.

3.21.B & 4.21.8 LIOUID EFFLUENTS 3.21.B.1 & 4.21.B.1 Concentration l

This specification is provided to ensure that the concentration of radioactive materials i

released in liquid waste effluents from the site to unrestricted areas will be less than the i

concentration levels specified in 10 CFR Part 20.1302. Thislimitationprovidesadditionall,,

assurance that the levels of radioactive materials in bodies of water outside the site will l

not result in exposures within (1) the Section IV.A guides on technical specifications in l

Appendix I,10 CFR Part 50, for an individual and (2) the limits of 10 CFR Part 20.1301 and 20.1302(b)(2)(1) to the population. The concentration limit for noble gases is based upon 4

the assumption that Xe-135 is the controlling radioisotope and its MFC in air (submersion) i

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was converted to an equivalent concentration in water using the methods described, in i

International Commission on Radiological Protection (ICRP) Publication 2.

Since SERVICE WATER is not a normal or expected source of significant radioactive release, ll routine sampli tion of 3 x 10'gr, and monitoring for radioactivity is precautionary. An activity concentra-i

^

pCi/ml in SERVICE WATER effluent is diluted in the discharge canal to about l

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1.5% of the 10 cm 20 Appendix B Table 2 Column 2 concentration with only one circulating 4

water pump operating. During normal Station operation the dilution would be even greater.

By monitoring SERVICE WATER effluent continuously for radioactivity and by confirmatory l sampling weekly, reasonable assurance that its activity concentration can be kept to a small 4

i fraction of the 10 C M Part 20.106 limit and within the Specification 3.21.B.2.a limit is i

provided.

By monitoring SERVICE WATER continuously and liquid radwaste continuously during discharge j

with the monitor set to alarm or trip before the limit specified in 10 CFR 20.1302 is j

exceeded, reasonable assurance of compliance with Specification 3.21.B.1.2 is provided.

Verification that radioactivity in liquid effluent averaged only a small fraction of the

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concentration limit is provided by calculations demonstrating compliance with Specifica-tion 3.21.B.2.a.

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Amendment No. 44,174

-216a21-4

3.21 & 4.21 RASES (Cont'd) 3.21.5 & 4.21.5 LIOUID EFFIDENTS (Cont'd) 3.21.B.1 & 4.21.B.1 Concentration (Cont'd)

Compliance with 10CFR Part 20.1302 impliesthattheconcentrationlimitrepresentedbyl 10Cm Part 20, Appendix B, Table 2 will be met within a suitable and reasonable averaging time for assessing compliance. That averaging time is dependent upon the resolving time of the measurements or estimates which are used to evaluate compliance. Assessment of compliance is done by sampling and analysis according to Specification 4.21.B.1.2, by estimating or measuring the maximum release flow and the minimum dilution flow coincident during the period of release represented by the sample, and by computing the concentration asafractionofthelimitintheUNRESTRICTEDareaperiodicallyonthebasisofthesel data.

3.21.B.2 & 4.21.B.2 Liauid Dose Specifications 3.21.B.2,

3. 21. C. 2 and 3. 21. C. 3 implement the requirements of 10 CPR Part 50.36a and of 10 CFR Part 50, Appendix I, Section IV.

These specifications state LIMITINGCONDITIONSFOROPERATION(LCO)tokeeplevelsofradioactivematerialsinLWRl effluents as low as is reasonably achievable. Compliance with these specifications will also keep average releases of radioactive material in effluents at small percentages of the limits specified in 10 CFR Part 20.1301. SurveillanceRequirementsprovideforthel j

measurement of releases and calculation of doses to verify compliance with the l

Specifications. Action statements in these Specifications implement the requirements of 10 CFR Part 50.36(c)(2) and 10 CFR Part 50, Appendix I, Section IV.A in the event an 140 i

is not met.

Annual dose limitations stated in Specifications 3.21.B.2, 3.21.C.2, and j

3.21.C.3 are not strict limits as used elsewhere in the Technical Specifications (are not an immediate safety concern) but do obligate NPPD to take the applicable reporting action required in Specifications 3.21.3.2.b, 3.21.C.2.b, or 3.21.C.3.b.

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10 CFR Part 50 contains two distinctly separate statements of requirements pertaining to effluents from nuclear power reactors. The first concerns a description of equipment to i

maintain control over radioactive materials in effluents, determination of design objectives, and means to be employed to keep radioactivity in effluents AIARA.

This requirement is stated in Part 50, Section 34a and Appendix I,Section II.

Appendix I,Section III stipulates that conformance with the guidance on design objectives be l

demonstrated by calculations (since demonstration is expected to be prospective).

The i

other is a requirement for developing LIMITING CONDITIONS FOR OPERATION in technical l specifications. It is stated in 10 CPR Part 50, Section 36a and Appendix I,Section IV.

f Both the intent of the Commission and the requirement are clearly stated in the opinion of the Commission;y relevant paragraphs from that document follow:

i c

Section 50.36a(b) of 10 CFR Part 50 provides that licensees shall be guided by certain considerations in establishing and implementing operating procedures specified in technical specifications which take into account the need for operating flexibility and at the same time ensure that the licensee will exert his best efforts to keep levels of radioactive materials in effluents as low as practicable.

The Appendix I that we adopt provides more specific guidance to licensees in this respect.

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i Amendment No. 49,174

-216a22-l I

s 3.21 & 4.21 BASES (Cont'd) l i

3.21.C & 4.21.C GASEGUS EFFLUENTS 3.21.C.1 & 4.21.C.1 concentration Specification 3.21.C.1.a is included to assure that a measure of control is provided over

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the concentration of radionuclides in air leaving the exclusion area. Radioactive noble games are monitored by instruments that provide a measure of release rate and cause i

automaticalarmwhenthenoblegasconcentrationOFFSITEisexpectedtoexceedthedosel rata specified in 3.21.C.1.a.

With prompt action to reduce the radioactive noble gas concentration in effluent following alarm initiation, it can be maintained at a small fraction of the annual limit. The specified release rate limits restrict the correspond-ing gamma and beta dose rates above background to an individual at or beyond the exclusion j

area boundary to $ 500 mrom/ year to the total body or to s 3000 mres/ year to th6 skin.

d j

Radioiodines and radiomeclides in particulate form are sampled with integrating samplers that permit assessment of the average release rate during each sample collection period.

l By complying with Specifications 3.21.C.2 and 3.21.C.3 the average OFFSITE concentration will be maintained at a small fraction of the 10 CFR Part 20.1302(b)(2)(1) concentration 4

limit.

3.21.C.2 & 4.21.C.2 Noble Gases Assessments of dose required by Specifications 4.21.C.2 and 4.21.C.3 to verify compliance with Appendix I,Section IV is based on measured radioactivity in gaseous effluent and i

on calculational methods stated in the ODAM.

Pathways of exposure and location of individuals are selected such that the dese to a nearby resident is unlikely to be l

'mderestimated. Dose assessment methodology described in the ODAM for gaseous affluent I

will be consistent with the methodology in Regulatory Guides 1.109 and 1.111. Cumulative i

and projected assessments of dose made during a quarter are based on historical average, or reference (the same period of record used in the design objective Appendix I evaluation) atmospheric conditions.

Assessments made for the annual radiological environmental report will be based on quarterly and annual averages, of atmospheric l

conditions during the period of release.

The bases for Specifications 3.21.C.2 and 4.21.C.2 are also discussed in the bases for l

Specifications 3.21.B.2 and 4.21.B.2.

j 3.21.C.3 & 4.21.C.3 Iodine and Particulates The bases for Specifications 3.21.C.3 and 4.21.C.3 are discussed in the bases for Specifications 3.21.B.2 and 4.21.B.2.

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l Amendment No. SS,174

-216a26-l L

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6.3 (Cont'd) i i

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A.

High Radiation Area r

In lieu of the " control device" or " alarm signal" required by paragraph 20.1601 of 10 CFR Part 20, each high radiation area in which the l

i deep dose equivalent in excess of 100 aren but less than 1000 area in one i

hour

• shall be barricaded (barricade will impede physical movement across l

the entrance or access to the high radiation area; i.e., doors, yellow and 4

4 magenta rope, turnstile) and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Special l

Work Permit (SWP)**.

Any individual or group of individuals permitted to j

enter such areas shall be provided with or accompanied by one or more of the l

l following:

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l 1.

A monitoring device which continuously indicates the radiation dose rate in the area.

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2.

A monitoring device which continuously integrates the radiation dose in l

the area and alarms when a preset integrated dose is received. Entry

(

l into such areas with this monitoring device may be made after the dose i

rates in the area have been established and personnel have been made knowledgeable of them.

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3.

A raciation protection qualified individual (i.e., qualified in radia-i tion protection procedures), with a dose rate monitoring device, who is i

responsible for providing positive control over the activities within l

the area and shall perform periodic dose rate monitoring at the fre-quency specified by Health Physics supervision.

In addition to the requirements of the above specification, areas accessible 1

to personnel with dose rates such that a major portion of the body could 1

receive in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a deep dose equivalent in excess of 1000 mres* shall be l

2 provided with locked doors to prevent unauthorized entry Doors shall remain locked except during periods of access by personnel under an approved SWP which shall specify the dose rates in the immediate work area. For l

individual high radiation areas accessible to personnel rhat are located j

within large areas, such as the containment, or areas V tre no enclosure exists for purposes of locking and no enclosure can be reasonably l

constructed around the individual areas, then that area shall be barricaded j

and conspicuously posted. Area radiation monitors that have been set to l

alarm if radiation levels increase, provide both a visual and an audible signal to alert personnel in the area of the increase. These monitors may be used to meet specification 6.3.A.1 provided that the dose rates and alarms have been established by radiation protection personnel.

Stay times j

or continuous surveillance, direct or remoto (such as use of closed circuit TV cameras), may be made by personnel qualified in radiation protection procedures to provide additional positive exposure control over the activities within the area.

• Measurement made at 12 inches from source of radiation.

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• • Radiation protection personnel or personnel escorted by radiation 2

protection personnel shall be exempt from the SWP issuance requirement j.

during the performance of their assigned duties, provided they are otherwise l

following plant radiation protection procedures for entry into high radiation areas, i

1 Amendment No. *2,*5,101,102,105,174

-226-2

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6.4.2.G (cont'd) f t

l usage evaluation per the ASME Soiler and Pressure Vessel Code Section

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III was performed for the conditions defined in the design specifica-l tion. The locations to be monitored shall be:

t a.

The feedwater nossles j

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b.

The shall at or naar the waterline c.

The flange studs 2.

Monitoring, Recording, Evaluating, and Reporting j

l a.

Operational transients that occur during plant operations will, at least annually, be reviewed and compared to the transient condi-l tions defined in the component stress report for the locations i

listed in 1 above, and used as a basis for the existing fatigue analysis.

l b.

The number of transients which are comparable to or more severe 3

than the transients evaluated in the stress report Code fatigue usage calculations will be recorded in an operating log book. For those transients which are more severe, available data, such as the metal and fluid temperatures, pressures, flow rates, and other j

conditions will be recorded in the log book.

c.

The number of transient events that exceed the design specifica-i tion quantity and the number of transient events with a severity t

greater than that included in the existing Code fatigue usage i

calculations shall be added. When this sua exceeds the predicated 8

number of design condition events by twenty-five, a fatigue usage l

evaluation of such events will be performed for the affected portion of the RCPB.

l H.

Records of current individual plant staff members showing qualifications and the completion of training.

l I.

Records for Environmental Qualification which are covered under the provi-sions of Specification 6.3.

l J.

Records of the service lives of all hydraulic and mechanical snubbers noted j

in 3.6.H.1, including the date at thich the service life commences and associated installation and maintenance records.

I i

2 j

See paragraph N-415.2, ASME Section III,1965 Edition.

2 The Code rules permit exclusion of twenty-five (25) stress cycles from secondary j

stress and fatigue usage evaluation.

(See paragraphs N-412(t)(3) and N-417.10(f) of the Summer 1968 Addenda to ASME Section III, 1968 Edition.)

1 l

i i

4 i

l 1.

Amendment No. % S2,92,1^",174

-229-

,r

1 a

6.5.1.C (Cont'd) l 1.

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mres/yr and their associated man rem exposure according to work and job functiona, e.g., reactor operations and l l

surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), vaste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measure-ments. Small exposures totaling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least i

80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

2.

A summary description of facility changes, tests or experiments in accordance with the requirements of 10CFR50.59(b). This report may be submitted annually or along with the Updated Safety Analysis Report (UFSAR) updates as required by 10CFR50.71(e).

3.

Documentation of all challenges to relief valves or safety valves.

D.

Hgnthly Operatine Report i

Routine reports of operating statistics, shutdown experience, and a narrative summary of operating experience relating to safe operation of the facility, shall be submitted on a monthly basis'to the individual designated in the current revision of Reg. Guide 10.1 no later than the tenth of each month fo11owin5 the calendar month covered by the report.

1 E.

Annual Radiolorical Environmental Reoort 1.

Routine radiological environmental reports covering the surveil-lance activities related to the Station operation during the previous calendar year shall be submitted to the NRC before May 1 of each year.

2.

The Annual Radiological Environmental Report shall include the l

following:

a.

A summary of doses to a MEMBER OF THE PUBLIC OFFSITE due to Cooper Nuclear Station aqueous.and airborne radioactive effluents, calculated in accordance with methods compatible with the ODAM.

b.

A summary of the results of the land use census required in Specification 4.21.F.2.

2 This tabulation supplements the requirements of $20.2206 of 10CFR Part 20.

l Amendment No. S,89,100,1':9,172,174

-231-