Text

Proposed Tech Specs,Revising LACBWR Tech Specs
	ML20147B779
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	02/22/1988
From:	; DAIRYLAND POWER COOPERATIVE
To:	;
Shared Package
ML20147B332	List: ML20147B327; ML20147B779;
References
	NUDOCS 8803020190
	Download: ML20147B779 (87)
	v • d • e

, . -_ - - -

\

J TABLE OF CONTENTS

, PAGE l .n SITE l-1

~1.1 -LOCATION 1 1.21 PRINCIPAL ACTIVITIES 1-1

2. , DEFINITIONS 2-1

3. APPLICABILITY -

3-1

- 4/5 PERFOINANCE REQUIREMENTS 4--I 4.1 FUEL STORAGE AND HANDLING - 1 4.151 General Fuel Handling and Storage 4-1 4.1.2 Fuel Element Storage Well 4-2 4.1.3 Fuel Element Storage Well Water Chemistry -

4-3

4.1.4 Fuel Element-Storage Well Water Supply. 4-4 4.2 CONTAINMENT' BUILDING 4-6

, 4.2.1 Containment Integrity- -

4-6 4.2.2 Containment Ventilation Dampers 4-9 4.2.3 Containment Vessel- , -- -- - -

4-10 4.2.4 Ventilation System Exhaust 11 4.3 ELECTRICAL POWER SYSTEMS-- 4-15

-4.3.1 A.C. Sources 4-15 4.3.2 Onsite A.C. Power Distribution Systems ---

-- 4-18 4.3.3 Onsite D.C. Power Distribution Systems-- --4-19 4.4 FIRE PROTECTION- -

23 4.4.1 Fire Detection Instrumentation 23 4.4.2 Fire Suppression Water System 25 4.4.3 Spray and/or Sprinkler Systems 28 4.4.4 Chemical Extinguishing - ---

4-29 4.4.5 Fire Ilose Stations- -

4-30 4.4.6 Yard Fire Hydrants and outside Ilose Houses ---

4-31 4.5 INSERVICE INSPECTION 4-33 4.5.1 Structural Integrity- -- 4-33 4.5.2 Pumps and Valves 4-34 i

8803020190 890222 PDR ADOCK 05000409 p PDR t- ,

w ,,

TABLE OF CONTENTS - ~(Cont'd) PAGE 4.6 RADIATION MONITORING --

4-36

'4.6.1 Radiation Monitors 36.

4.6.2 Post-Accident Radiation Monitoring and-Instrumentation--4-37 4.6.3 Sealed Source Contamination 4-38

- 4.7 RADIOACTIVE EFFLUENTS 41 4.7.1 Radioactive Effluents 4-41 4.7.1.1 Instrumentation- 4-41 4.7.1.2 Concentration 4-43 4.7.1.3 Dose- 4-43 4.7.2 Radioactive Gaseous Effluent 4-46 4.7.2.1 Instrumentation 4-46 4.7.2.2 Instantaneous Dose Rate 4-49 4.7.2.3 : Dose, Noble Gas 4-51 4.7.2.4 Dose, Radionuclides Other Than Noble Gas 4-52 4.7.3 Solid Radioactive Waste- 4-53 4.7.4 ~ Total Dose ---

4-54 4.8 ilADIOLOGICAL ENVIRONMENTAL MONITORING AND 4-55 INTERLABORATORY COMPARISON

6. ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6-1 6.0 ORGANIZATION- 6-1 6.3 FACILITY STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND AUDIT- 6-3

'6.5.1 Operations Review Committee (ORC) ------

G-3 6.5.2 -. Safety Review Committee (SRC) 6-3 6.6 PROCEDURES 8 6.7 CONTROL OF MAINTENANCE AND TESTING ACTIVITIES 6-10 6.8 REPORTING REQUIREMENTS-- --

6-10 6.8.1 Routine Reports 10 6.8.2 Special Reports -G-12 6.8.3 Licensee Event Reports-- C-12 6.8.4 Immediate Notification Requirements =6-13 L 11 o

- - ,--cy-re .-.--ege nm,e-r--,p re--w-,--- -.-r+ .-,---.mr.

m--

TABLE OF CONTENTS - (Cont'd) PAGE CU 6.9 RECORD RETENTION 6-13 6.10 RADIATION PROTECTION PROGRAM 14 6.11 IIIGli RADIATION AREA 6-14 e

i11 !

. - ~

I

( s gn

. Q

-1. SITE

======================================u==========================

1.1 I4 CATION The La Crosse Boiling Water Reactor (LACBWR) is located on the east ba20c of the Mississippi River, approximately 19 miles south of the city of La Crosse, Wisconsin, and 1 mile south of the Village of Genoa, Wisconsin.

1.2 ERp"'IPAL ACTIVITIES The principal' activity carried on at the La Crosse Boiling Water Reactor .

shall be possession of the facility. The major activities shall include the monitoring of the reactor and plant equipment, storage and . handling of the .

- reactor fuel, maintenance of systems required for safe storage activities, monitoring effluents,'and analyzing environmental samples to assure the health and safety of .the public.

1 l

l-1

2. DEFINITIONS

=================================================================

The following terms are defined so that uniform interpretation of these specifications may be achieved. When these terms appear in capitalized type, the following definitions apply in these Technical Specifications.

ACTION ACTION shall be that part of a spncification whicb prescr5bes remedial measures required under designated cenditions.

CIIANNEL CALIBRATION A CilANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CIIANNEL CALIBRATION shall encompass the entire chennel including the sensor and alarm and/or trip functions, and shall include the CIIANNEL FUNCTIONAL TEST. The CllANhTL car.IBRATICN may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL ClfECK A CHANNEL CIIECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels neasuring the same parameter.

CHANNEL FUNCTIONAL TEST A CllANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b. Bistable channels - the injection of a real or simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

MNTAINMENT INTEGRITY CONTAINMENT INTEGRIrr shaj' exist when:

a. All penetrations required to be isolated are either:

1. Capable of being closea by an OPERABLE containment automatic isolation valvo system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position.

2-1

s DEFINITIONS - (Cont'd)

=================================================================

b. The freight door is closed,

c. Each air lock is OPERABLE,

d. The containment leakage rates are within the limit, and

e. The sealing mechanism associated with each penetration (e.g.,

welds, bellows, o-rings) is OPERABLE.

EFFLUENT RELEASE BOUNDARY The Dairyland Power Cooperative property line within the 1109 ft. (338m) radius EXCLUSION AREA is the EFFLUENT RELEASE BOUNDARY. See Figure 4/5.7.

EXCLUSION _ AREA The EXCLUSION AREA is defined as the area within an 1109 ft. (338m) radius from the centerline of the Cantainment Building. This was the area established per 10 CFR 100 as the EXCLUSION AREA for plant siting and operation.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the peri'ormance of Surveillance Requirements shall correspond to the intervals defined in Table of Surveillance Frequency Notation.

FU3L HANDLING FUEL HANDLING shall be the movement of any irradiated fuel within the Containment Building, Suspension of FUEL HANoLING shall not preclude completion of mosement of the fuel to a safe conservative position.

MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include any person who is not occupationally associated with the utility. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or make deliveries. This category does include persons who use portions of the site for recreational or other purposes not associated with the utility.

OFFSITE DOSE CAlfULATION MANUAL (ODCM)

An OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be a manual containing the methodology and parameters to be used for the calculation of offsite doses due to radioactive gaseous and liquid effluents and for the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints.

It shall describe the radiological environmental monitoring program.

2-2

I cr DEFINITIONS -(Cont'd).

=================================================================

OPERABLE-OPERABILITY A system,' subsystem,-train, component or device shall be OPERABLE or have j OPERABILITY when it is capable of performing its specified function (s) and

.when all necessary attendant instrumentation, controls, a normal or an.

emergency electrical power source, cooling or seal water, lubrication or E other auxiliary equipment that_are required for the system, subsystem, train, h

component or device to perform its function (s) are also capable of performing

( their related support function (s).

r PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM shall contain the current formula, sampling,

-analyses, tests, and determinations to be made to ensure that the protissing and packaging of solid radioactive wastes based on demonstrated procest \ng of actual or simulated solid wastes will be accomplished in such a way as so assure compliance with 10 CFR Part 20,10 CFR Part 61 and Federal and State regulations and other requirements governing the transportation and disposal of the radioactive waste.

REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 6.8.3. A Licensee Event Report shall be submitted for REPORTABLE EVENTS.

RESTRICTED AREA A RESTRICTED AREA shall be any area within the EXCLUSION AREA, access to which is controlled by the licensee for purposes of protection of individuals from exposure to ionizing radiation and radioactive materials.

SOLIDIFICATION SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive cource.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, sub.: stems, trains or other v designated components obtained by dividing the specified test interval into n equal subintervals,

b. The testing of one system, subsystem, train or oth > designated component at the beginning of each subinterval.

2-3 '

I I

i J

DEFINITIONS - (Cont'i.)

=___________________g_==____=====u====_==____,=__=__=___ =__=,_==__=__=_____=.

U UNRT3TRICTED AREA' AnLUNRESTRICTED AREA shall be any area not controlled.by the licensee for purposes of protection af MEMBERS OF THE PUBLIC from. exposure 'to ionizing radiation and radioactive materials.

'3URVEILLANCE FREQUENCY NOTATION NOTATION FREQUENCY S- At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 6 months.

A At least.once per 12 months.

R ht least once por 18 months.

P Completed prior to each use or release.

N.A. Not applicable.

\

)

2-4 1

3. APPLICABILITY LIMITING CONDITION FOR OPERATION

t ===================================================================

3.1 Limiting Conditions for Operation and ACTION requirements shall be applicable during the specified applicable condition for each specification.

3.2 Adherence to the requirements of the Limiting Condition for Operation and/or associated ACTION within the specified time interval shall constitute compliance with the specification. In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval completion of the ACT7'N statement is not required.

3.3 Entry into specified applicability state shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted.

SURVEILLANCE REQUIREMENTS

=================================================================

3.4 Surveillance Requirements shall be applicable during the specified applicable conditions for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

3.5 Each Surveillance Requirement shall be performed within the specified '

time interval with:

a. A maximum allowable extension not to exceed 25% of the surveillance interval,

b. A total maximum combined interval time for any 3 consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.

3.6 Performance of a Surveillance Requirement within the specified time interval shall constitute compliance with OPERABILITY requirements for a Limiting Condition for Operation and associated ACTION statements unless otherwise required by the specification. Surveillance requirements do not have to be performed on inoperable equipment or on equipment not required to be OPERAB12.

! 3.7 Entry into a specified applicable condition shall not be made unless the i Surveillance Requirement (s) associated with the Limiting Condition for l Operetion have been performed within the stated surveillance interval or as otherwise specified.

L 3-1

APPLICABILITY BASES

=================================================================

The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section 4/5.

3.1 This specification defines the applicability of each specification in terms of specified applicability conditions and is provided to delineate specifically when each specification is applicable.

3.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting condition for Operation and associated ACTION requirement.

3.3 This specification provides that entry into specified applicability condition must be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions contained in the ACTION statements.

Ine intent of this provision is to ensure that activities are not initiated with either required equipment or systems inoperable or other specified limits being exceeded.

Exceptions to this provision have been provided for a limited number of specifications when performance of activities with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications.

3.4 This specification provides that surveillance activities necessary to ensure that the Limiting Conditions for Operation are met and will be performed during the specified applicability conditions for which the Limiting Conditions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the specified applicability conditions are provided in the individual Surveillance Requirements.

3.5 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide flexibility because of scheduling and performance consideration. The phrase "at least" associated with a surveillance frequency requirement does not negate these allowable tolerances for performing surveillance activities; instead it permits more frequent performance of surveillance activities than required by the specification.

The tolerance values, taken eithen individually or consecutively over three !

test intervals, are sufficiently restrictive to ensure that the reliability associated with the snivei.11ance activity is not significantly degraded beyond that obtained fra. the nominal specified interval.

l 3-2 l

m --

-- g-APPLICABILITY B ASES '- (Cont'd) -

~

3.6' The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting Condition for Operation. Under this criteria, equipment, systems or components are assumed to be OPERABLE if the associated surveillance

< activities have been satisfactorily performed within'the specified time

-interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items'are found or known to be inoperable although still meeting the Surveillance Requirements.

3.7 This specification ensures that the surveillance activities associated with a Limiting Condition for Operation have been performed-within the

- specified time interval prior to entry into specified applicability condition. The intent of this provision is to ensure that survel'. lance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting condition for Operation.

1 ll a 3-3 h

4. PERFOINANCR REQUIREMENTS l

4/S.1 [UEL STORAGE AND ilANDLING 4.1.1 General Fuel Storage and Handling Requirements 4.1.1.1 Irradiated fuel elements shall be stored underwater in spent fuel storage racks that are positioned on the bottom of the Fuel Element Storage Well, in approved onsite dry storage containers 3 or in an approved shipping cask.

4.1.1.2 During the handling of irradiated fuel elements that have been operated at power levels greater than 1 Mwt, the depth of water in the reactor upper cavity and/or the Fuel Element Storage Well shall be at least 2 feet above the active fuel.

4.1.1.3 With the exception of a spent fuel shipping cask, the core spray bundle, the transfer canal shield plug and the other components and fixtures that are normally located and used within the storags well, no objects heavier than a fuel assembly shall be handled over the Fuel Element Storage Well, l

4-1

FUEL STORAGE AND HANDLING ,

I

_F,URL ELEMINT S10 RAGE WELL i LIMITING CONDITION IOR OPERATION !

============================================================================ ;

4'.1. 2 The Fuel-Element Storage Well (FESW) shall meet the following requirements

I

a. The Fuel Element Storage Well water level shall'be at least 16 feet above any irradiated fuel stored in the spent fuel storage racks, and

b. Water in the storage well shall be maintained at a temperature

< 150oF.

APPLICABILITY: At all times.

ACTION:

l l

st . With water level'less than 16 feet above any irradiated fuel stored in }

the Fuel Element Storage Well storage racks, take immediate action to !

restore water level and suspend all operations involving FUEL HANDLING. j

b. With water temperature in the storage well above 1500F, take actions to i reduce water temperature to .{ 1500F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and suspend any i evolutions involving rTEL HANDLING.

SURVEILLANCE REOUIREMENTS 4

============================================================================= .

5.1.2.1 The Fuel Element Storage Well water level and FESW System water temperature shall be monitored at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . !

5.1.2.2 The Fuel Element Storage Well water level indication channel shall be calibrated at least once per 18 months. ,

N l

e T t I

e 4-2 ,

h

?

FUEL STORAGE AND HANDLING FUEL ELEMENT STORAGE WELL WATER CHEMISTRY LIMITING CONDITION FOR OPERATION

==_=============================================================

4.1.3 The Fuel Element Storage Well water shall meet the following requirements:

a. Chloride concentration- 1 0.5 ppm

b. pH -

5.3 - 8.6

c. Conductivity 1 10 unho/cm

d. Gross beta-ga==a activity 1 0.1 uCi/s1 APPLIC_ ABILITY: Whenever fuel is stored in the Fuel Element Storage Well.

ACTION:

With the Fuel Element Storage Well water chemistry or radiochemistry limits exceeded, ini'. tate action to restore water quality to within the limits. t SURVEILLANCE REQUIRDIENTS

================================================================

5.1.3 The Fuel Element Storage Well chemistry and radiochemistry parameters specified in this section shall be determined at least once every 7 days.

9 I

4-3 r

FUEL STORAGE AND HANDLING FUEL ELEMENT _ STORAGE WELL WATER SUPPLY LIMITING CONDITION FOR OPERATION

============================================================================ t 4.1. 4 At least one of the following Fuel Element Storage Well water supplies shall be OPERABLE:

a. The Demineralized Water Tank with a minimum water level of two feet, (approximately 5000 gallons) or

b. The Overhead Storage Tank with a minimum contained water volume of 5,000 gallons.

APPLICABILITY: At all times.

ACTION:

With neither the Domineralized Water Tank nor the Overhead Storage Tank OPERABLE, restore level in at least one tank within 7 days.

SURVEILLANCE REQUIREMENTS

================================================================

5.1.4.1 The Demineralized Water Tank shall be demonstrated OPERABLE by verifying the minimum water level in the tank at least once per 7 days.

5.1.4.2 The Overhead Storage Tank shall be demonstrated OPERABLE by verifying the minimum contained water volume at least once per 7 days, i

4-4 i

t BASES I

=====================xz=========================================

4/5.1 FUEL STORAGE AND HANDLING Spent fuel storage is provided in the spent fuel storage racks located in the Fuel Element Storage Well'within the Containment Building. The spent fuel '

storage rocks are designed with a nominal 7.0 inch center-to-center distance between fuel assemblies in each individual rack assembly with a boron containing poison slab between each storage location to ensure Lte of f'0.96 l i when flooded with unborated water. Fuel stored in the storage well is restricted to fuel with stainless steel cladding which has a U-235 loading of f 22.6 grams per axial centimeter. A fuel handling system'is provided which is capable of remotely handling fuel assemblies one at a time.

The Fuel Element Storage Well System is capable of maintaining water temperature f 150*F and maintaining water quality within the established ,

limits.

Minimum water coverage limits above fuel stored in the storage racks and fuel f being handled have been established to provide adequate shielding to protect personnel and to provide adequate cooling. Limits to the handling of heavy !

objects over the Fuel Element Storage Well have been established to reduce the probability of a heavy load drop into the storage well.

A minimum Fuel Element Storage Well water supply has been established to '

ensure that sufficient water is available onsite to provide short-teris makeup to the storage well while an alternate supply is being established if needed. !

4-5

"w

'4/5.2 CONTAINMENT BUILDING CONTAIPNENT INTEGRITY LIMITING CONDITION FOR OPERATION

========t=u==2==================================================

4.2.1 CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: During FUEL HANDLING.

ACTION:

If CONTAINMENT INTEGRITY does not exist, suspend all FUEL llANDLING within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

SURVEILLANCE REQUIREMENTS

====================================== =========================

5. 2.1.1 Containment Building'gasketed closures and ventilation system closures which have been subjected to maintenance, repair or other operations which might affect their performance shall, before any subsequent operation for which containment integrity is required, be tested for leak tightness using the soap-bubble technique (or other method of equivalent sensitivity).

This test shall be performed using a pressure differential no less than 0.5 psi and the results shall be used as a guide in evaluating leakage.

5.2.1.2 Any new or existing containment penetrations which are placed in or removed from service shall be tested for leak tightness using the soap-bubble technique (or other method of equal sensitivity) at a pressure differential of at least 10 psi. The test shall be performed before any subsequent FUEL HANDLING is conducted.

5.2.1.3 The door seals on the containment personnel and emergency airlocks will be visually inspected for degradation prior to TUEL HANDLING if not inspected within the preceding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

5.2.1.4 The door seals on the containment personnel and emergency airlocks will be replaced periodically in accordance with manufacturer's recommendations.

4-6 u_

CONTAINMENT BUILDING CONTAINMENT INTEGRITY SURVEILLANCE REQUIREMENTS - (Cont'd) 5.2.1.5 Individual leak detection tests shall be performed on Containment Isolation Valves and the following penetrations at a pressure of at least 10 psig prior to FUEL IIANDLING if not performed within the preceding 90 days, except for the freight door. The freight door shall be tested following each closure prior to FUEL IIANDLING. The containment penetrations shall be tested using the soap-bubble technique (or other methods of equivalent sensitivity) or by determining the rate of pressure loss of pneumatically pressurized test chambers. The combined leakage rate for all penetrations and valves tested shall be no greater than 50 SCFl!. The penetrations to be tested are:

o the electrical penetrations,

' o the reactor building spray valve shaft penetration, o the freight door, o the main steam line penetration, o the feedwater line penetration, o the heating steam line and condensate return penetration, o the containment building airlocks, and a the flanges of the e-ntilation system inlet and exhaust ducts.

5.2.1.6 Containment Isolation Calves shall be tested prior to FUEL HANDLING if not tested within the precedi.'g 60 days to demonstrate automatic closure on the conditions specified in Tr 21e 4.2.1 unless the Containment Isolation Valve is deactivated in its closo! position or the penetration is isolated by at least one manual valve or blind flange.

4-7

1 CONTAINMENT BUILDING CONTAI!NENT INTEGRITY TABLE 4.2.1 CONTAINMENT ISOLATION' VALVES CIIANNEL, SENSOR, VALVE OR CONDITION SETPOINT Containment Ventilation Inlet and Outlet A C Dampers B D Loss of Control Air NA Loss of Electrical Power Supply NA Containment Vent Header Valve A C B D Decay Heat System Blowdown Valve A C B D Heating Steam condensate Return Valve A C B D Retention Tank Pump Discharge Valve A C B D TABLE NOTATION:

A. Reactor Containment Building Ventilation Particulate or Gaseous Activity Monitor B. Fuel Clement Storage Well Level C. I Radiation levels which Correspond to the limits of Specification 4.7.2.2.

D. 1 690' MSL 4-8 o

CONTAINMENT BUILDING CONTAINMENT VENTILATION DAMPERS LIMITING CONDITION FOR OPERATION

=================================================================

4.2.2 The Containment ventilation inlet and outlet dampers shall be OPERABIR with isolation times of less than or equal to 10 seconds.

APPLICABILITY: Whenever CONTAIN4ENT INTEGRITY (Specification 4.2.1) is required.

ACTION:

l l a. With one or more of the above ventilation damper (r.) inoperable, suspend FUEL HANDLING or isolate the affected penetration, with a deactivated automatic valve secured in its closed position or with a blind flange, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

b. The provisions of Specification 3.3 are not applicable if the affected penetration is isolated.

SURVEILLANCE REQUIREMENTS

=================================================================

5.2.2.1 The ventilation dampers shall be demonstrated OPERABLE prior to returning the damper to service after maintenance, repair or replacement work is performed on the damper or its associated actuator, control, or power circuit by performance of a cycling test, and verification of isolation time.

5.2.2.2 The isolation time of each above damper shall be determined to be within its limit when tested pursuant to Specification 5.2.1.6.

5.2.2.3 The seat rings of the ventilation inlet and outlet dampers shall be replaced at least once per 5 years.

5.2.2.4 The Containment ventilation system dampers shall be subjected to quarterly leakage tests, in addition to the tests required by Specification 5.2.1.5. The tests shall be conducted at an initial pressure of at least 10 psig. Excessive degradation is determined not to exist and the isolation valve (s) is considered operable if pressure decreases by less than 5 psi in a ten minute test period.

4-9

4

n. ,

CONTAI MtNT BUILDING CONTA!!9fENT VESSEL LIi41 TING CONDITION FOR OPERATION

=================================================================

4.2.3 The Containment' Vessel shall be protected by the follcaing

. requirements:

a. At least one Containment Building vacuum breaker shall be set to relieve at a differential (external-over-internal) pressure not exceeding 0.5 psi.

b. -The Containment Building steel shell temperature shall be greater than OoF when the Containment Building pressure exceeds 10.4 psig.

APPLICABILITY: At all times.

ACTION:

a. With no vacuum breakers OPERABLE, restore at least one vacuum breaker to service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. With the Containment Building steel shell temperature less than OoF with pressure exceeding 10.4 psig, restore the temperature'and/or pressure to -

within the limits within 30 minutes and perform an engineering ;

determination of the effects of the out-of-limit condition on the fracture toughness properties of the Containment Building. I SURVEILLANCE REQUIREMENTS

================================================================-

5.2.3.1 The Containment Building vacuum breakers chall be tested for proper operation at least once every 18 months.

5.2.3.2 containment Building steel shell temperature shall be monitored every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when Containment Building pressure exceeds 10 psig.

r F

P f

s

+

?

l l i l

i t ,

l 4-10 j r

I. m

(

l

e 9. , - . - 3, . .

Y sis

~

CONTAIl#ENT BUILDING

' VENTILATION SYSTEM EXHAUST LIMITING CONDITION FOR OPERATION -

==============================================e=================r.=====u===== ,

t 4.2.4' The Containment Building Ventilation System exhaust shall be through l particulate filters.

APPLICABILITY: Whenever the Containment Ventilation outlet dampers are open.

' ACTION:- l With Containment Building Ventilation System exhaust being discharged without filtration, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report which discusses the circumstances and

t. hat' action will be taken to prevent a recurrence. ,

SURVEILLANCE REQUIREMENTS

=================================================================

5.2.4 Cumulative dose contributions from gaseous releases are calculated per Sections 5.7.2.3 and 5.7.2.4. j i

f i

f I

+

I I

'e 9

1; I

r P

I t

4-11 !

t

_'Y

s 4/5.2. CONTAINMENT DUILDING BASES

=============================r.============================.======

4/5.2.1 Containment Integrity The containment vessel was crisinally designed to be capable of containing an internal pressure of 52 psig at 280oF. The established containment leakage rate corresponded to a leakage rate that, per 24-hour period, did not exceed 0.1 percent by weight of a steam-air mixture at 273o F, 28.5 psig, and a steam-to-initial-air ratio of 2.2. Leakage testing was conducted at 52 psig; the maximum allowable containment leakage was 50 SCFil.

With the plant in the SAFSTOR condition, there is no longer a postulated accident that would result in containment pressurization. The requirement for CONTAINMENT INTEGRITY during fuel handling has been retained. Leakage testing is conduct)d at a lower pressure, since there would be no pressure head during any postulated accident.

Containment isolation signals are based on high Containment Building ventilation activity and low Fuel Element Storage Well level since the irradiated fuel is stored in the storage well in the Containment Building and these conditions can be indicative of a postulated accident.

The design bases for containment vessel penetrations remains as origiaally established.

Personnel and Emergency Airlocks: The airlock doors shall be held closed by locking bars or locking rings which clamp against locking lugs on the doors, and these doors shall be sealed by means of gaskets. The doors of each airlock shall be mechanically interlocked so that one door must be completely closed and sealed before the other door can be opened. The door operation shafts shall be sealed by mechanical seals at both the inner and outer door frames.

Freight Door: The freight door shall be sealed by means of gaskets and shall be held closed by means of hinged bolts and lugs.

Electrical Penetrations: Each cable penetration shall have two gas-tight seals to prevent leakage from the containment vessel. Test chambers shall be provided to permit pressurized leak testing between the seals. Closures provided for electrical penetration chambers shall be welded or shall be bolted and sealed by gaskets.

Mechanical Penetrations: The handwheel shaft for the building spray system shall be provided with two gas-tight seals to prevent leakage from the containment vessel.

4-12

[

CONTAINMENT BUILDING BASES - (Cont'd)

Piping and Ventilation Duct Penetrations: Piping and duct penetrations shall be sealed by means of welding directly to the containment vessel wall or by means of corrugated-type expansion joints which shall be welded to the pipe at one end and to a containment vessel nozzle at the other end. All piping ,

and duct penetrations, other than spares which have welded pipe caps, shall conform to the following:

a. Two valves in series shall be provided at containment penetrations for piping connected directly to the primary system or for piping, other than vacuum breaker lines which is open both to the building atmosphere and to the outside atmosphere. At least one of these valves shall be capable of closing automatically, unless one valve is normally closed during plant operation. The second valve shall be operable either from the control room or from some other location which would be accessible after an accident,

b. Two dampers shall be provided on each of the containment Building ventilation penetrations. The dampers shall be capable of being closed automatically.

Testing is required following any placement of new penetrations in service or after any work on an existing penetration that might affect its sealing capability prior to FUEL HANDLING.

4/5.2.2 Containment Ventilation Dampers The Nuclear Regulatory Commission requested that similar Technical Specifications per Generic Item B-24 and NUREG-0737, Item II.E.4.2, be submitted to help assure operability of containment ventilation dampers. The ACTION statement has been modified due to the plant's permanent shutdowr, DPC had previously committed to replace the containment ventilation inlet and outlet dampers' resilient sealing material at least once each 5 years until such time as additional in-situ data can be accumulated to justify a longer interval. If in-situ data is accumulated which supports a longer seal replacement interval, a change to Specification 5.?.2.3 may be requested.

The additional leakage test was established to detect any gross degradation of the dampers.

4/5.2.3 Containment Vessel The containment vessel was designed to the requirements of the ASME Boiler and Pressure Vessel Code,Section VIII (1962 Edition) as modified by the nuclear code cases applicable as of June 1962. Tne vessel plate is ASTM A201, Grade B steel, conforming to the test requirements of ASTM A300. The temperature / pressure limit was established to assure the fracture toughness properties of the vessel are maintained. The containment vessel is capable of withstanding an external-over-internal pressure of 0.5 psi. Therefore, a vacuus relief system is provided which is capable of limiting the external-over-internal pressure to 0.5 psi.

4-13

r CONTAINMENT BUILDING BASES - (Cont'd) 4/5.2.4 Ventilation System Exhaust Filtration of the contaitunent Building Ventilation System exhaust 42 required to reduce the amount of radioactive particulatos being released to the environment. This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50.

4-14

4/5.3 ELECTRICAL POWER SYSTEMS A.C. SOURCES LIMITING CONDITION FOR OPERATION

=================================================================

4.3.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

.a. The physically independent circuit between the offsite transmission network and the onsite power distribution system and at least one electrical feeder from 2400-volt bus 1A or 1B energizing either, but not both, essential switchgear buses,

b. Diesel Generator lA or IB, aligned to, but not energizing, the same essential switchgear bus, with:

1. A day tank containing a minimum of 80 gallons of fuel for diesel generator lA or 170 gallons of fuel for Diesel Generator 1B.

2. A fuel storage system containing a minimum of 200 gallons of fuel for Diesel Generator LA or 2500 gallons of fuel for Diesel Generator IB.

3. A fuel transfer pump.

APPLICABILITY: At all times.

ACTION:

With less than the above required A.C. electrical power sources OPERABLE, suspend all operations involving FUEL HANDLING until at least the minimum required power sources are OPERABLE.

SURVEILLANCE REQUIREMENTS

=================================================================

5.3.1.1 The physically independent circuit between the offsite transmission network and the onsite power distribution system and the electrical feeds from 480-volt bus lA and IB to the associated essential switchgear bus shall be determined OPERABLE at least once per 7 days by verifying correct breaker alignments and indicated power availability.

4-15

l l

RLBCTRICAL' POWER SYSTEMS' I A.C. SOURCES i SURVEILLANCE REQUIREMENTS - (Cont'd)- ,

.$.3.1.2 Each diesel generator shall be demonstreted OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1. Verifying the fuel level in the day fuel tank. ,

2. Verifying the fuel level in the fuel storage tank.

3. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the day tank.

4. Verifying the diesel generutor starts from ambient condition, and reaches rated bus voltage in less than or equal to 20 seconds.

5. Verifying tho diesel generator is loaded with the design test load and operates for greater than or equal to 60 minutes.

6. Verifying the diesel generator is aligned to provide emergency power-to the associated essential buses.

b. At least once per 92 days by verifying that a sample of diesel fuel from within 3 inches of the bottom of each fuel storage tank, is within the acceptable limits specified in T.'a ie 1 of ASTM D975-81 when checked for water and sediment, and for viscosity.

c. At least once per 18 months by:

1. Subjecting the diesel to an inspection in accordance with its manufacturer's recommendations for this class of standby service.

2. Verifying the diesel generator capability to reject a load greater than or equal to 41 kw load on Diesel Generator lA and l greater than or equal to 50 kw on Diesel Generator IB without l tripping.

l

3. Simulating a loss of offsite power and:

a) Verifying de-energization of the essential switchgear buses associated with Diesel Generators lA end IB.

b) Verifying the diesel generator starts **com ambient condition on the auto-start signal, energizes the essential switchgear bus with permanently connected loads, and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.

4-16

1m

.[ . "

, >0

r N n-4 , ,

~ ELECTRICAL' POWER SYSTEMS

' A.C. SOlmCES

SURVEILLANCE REQUIREMENTS - (Cont'd).

oc)_ Verifying'that~all Diesel Generator IB trips, except engine overspeed, overcrank, and~ generator-differential, are automatically bypana ed upon loss of voltage on.the essential bus.~

n '

E ~4. Verifying that' all available loads to each diesel generator-do not exceed the continuous' load rating of 250 kw for Diesel Generator lA and 400 kw for Diesel Generator IB.-

5 . 3 . 1 . 31 The starting and control power battery and battery charger of each y

. diesel generator shall'be demonstrated OPERABLE: "

a. At least'once per-7 days by verifying that: ;

1. .The electrolyte level of each battery is above the plates, ,

L

2. The pilot cell specific gravity, corrected to 770F and normal electrolyte level, is greater than or equal to 1.180, and has -

' not decreased more than 0.04 from the value obr.erved during the :

_ previous test, and

3. The overall battery voltage is greater than or equal to 24 volta for Diesel Generator IA and greater than or equal to 32 volts for Diesel Generator IB.-

b. At least once per 18 months by verifying that:

1. The batteries, cell plates, and battery racks show no visual ,

indication of physical damage or abnormal dit crioration, and ,

2. The battery-to-battery terminal connect.vas are clean, tight, "

' free of corrosion and costed with anti-corrosion material, j i

I h

I i

4-17 r

A- .-

.e ,

\l

.RLECTRICAL POWER SYSTEMS ONSITE A.C. POWER DISTRIBUTION SYSTEMS

. LIMITING CONDITION FOR OPERATION-

==================================== ========================================-

'4.3.2 As a minimum, the following A.C. electrical ~ buses shall be OPERABLE and energized from sources of power other than.a diesel generator but aligned to an OPERABLE. diesel generator.

a. 480-Volt Essential Switchgear Bus lA or 480-Volt Diesel Building Essential Switchgear Bus 1B.

b. 120-Volt A.C. Non-Interruptible Bus IB.

c. 120-Volt Turbine Building Regulated Bus.

APPLICABILITY: At all times.

ACTION:

With'less than the above required A.C. distribution system buses OPERABLE, suspend all-operations involving FUEL HANDLING within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS

===========================-=====================================

5.3.2 Each of the above required A.C. distribution system buses shall be determined OPERABLE at least once per 7 days by verifying correct breaker alignment and indicated power availability.

4-18

(

L

ELECTRICAL POWER SYSTEMS ONSITE D.C. POWER DISTRIBUTION SYSTEMS LIMITING CONDITIONS FOR OPERATION

=================================================================

4.3.3 The following D.C. buses shall be. energized and OPERABLE:

a. Generator Plant 125-volt D.C. bus, a full capacity charger, and a 125-volt battery bank, and

b. Diesel Building 125-volt D.C. bus, a full capacity charger, and a 125-volt battery bank or the Reactor Plant 125-volt D.C. bus, a full capacity charger, and a 125-volt battery bank.

APPLICABILITY: At all times.

ACTION:

With less than the above required 125-volt DC buses OPERABLE, suspend all operations involving FUEL llANDLING within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS

=================================================================

5.3.3.1 Each of the above required D.C. distribution system buses shall be determined OPERABLE at least once per 7 days by verifying correct breaker alignment and indicated power availability.

5.3.3.2 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1. The electrolyte level of each pilot cell is between the minimum and maximum level indication marks,

2. The pilot cell specific gravity, corrected to 770F and normal electrolyte level, is greater than or equal to 1.200,

3. The pilot cell voltage is greater than or equal to 2.0 volts, and

4. The overall battery voltage is greater than or equal to 120 volts.

4-19

RLECTRICAL POWER SYSTEMS ONSITE D.C. i W ER DISTRIBUTION SYSTEMS SURVEILLANCE REQUIREMENTS - (Cont'd)

b. At least once per 92 days by verifying that:

1. The voltage of each connected cell is greater than or equal to 2.0 volts under float charge and has not decreased more than 0.3 volts from the value observed during the original acceptance test,

2. The specific gravity, corrected to 770F, of each connected cell is greater than or equal to 1.200 and has not decreased more than 0.04 from the value observed during the previous test, and

3. The electrolyte level of each connected cell is between the minimum and maximum level indication marks.

c. At least once per 18 months by verifying that:

1. The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration.

2. The cell-to-cell and tenninal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

3. The battery capacity is adequate to supply and maintain in OPERADLE status a load equal to or greater thun all of the emergency loads or all of the actual e:Sc.rgency loads for 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months when the battery is subjected to a battery service test.

4. At the completion of the above test, the battery charger shall be demonstrated capable of recharging its battery while supplying normal D.C. loads. The battery shall be charged to at least 90% capacity in less than or equal to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4-20 L

f~

[

BASES 3======================================================================== ===

4/5.3 ELECTRICAL POWER SYSTEMS The OPERABILITY of the normal and backup electrical power systems ensures that sufficient power will be available to supply necessary plant equipment normally and following a loss of offsite power event.

Normally, all power is supplied by a connection to the external grid via the reserve auxiliary transformer.

The onsite power distribution system is divided between two independent 2400 volt buses and each component of duplicate equipment is supplied by a different bus. Each 2400 volt bus also supplies power to a 480-volt bus. If the supply to either 480 volt bus fails, a connection may be manually made from the other 480 volt bus using a control switch in the control room.

Duplicate equipment is also supplied by the 480 volt buses.

Each 480 volt bus normally supplies power to its associated essential bus.

The 480 Volt Bus lA supplies the 480 Volt Essential Switchgear Bus lA. The 480 Volt Essential Switchgear Bus lA supplies Turbine Building Motor Control Center lA which supplies the Turbine Building 120 Volt Bus which in turn supplies the Turbine Building 120 Volt Regulated Bus. Diesel Generator lA serves as the backup power supply for the 480 Volt Essential Switchgear Bus lA.

Turbine Building Motor Control Center IA supplies power for the Reactor Plant Battery Charger. The Reactor Plant Battery Charger carries the load and charges the batteries on the Reactor Plant 125 Volt DC Bus. The Reactor Plant 125 Volt DC Bus normally supplies power for Static Inverter lA which supplies power to instrumentation and control circuits connected to 120 Volt AC Non-Interruptible Bus lA. In case of Static Inverter lA trouble, a transfer switch located in Static Inverter lA automatically transfers to its alternate power supply which is Turbine Building 120 Volt Regulated Bus.

The 480 Volt Bus 1B supplies power to the 480 Volt Turbine Building Motor Control Center ID. Turbine Building Motor Control Center ID is the normal supply for the Generator Plant 125 Volt DC Bus through the Generator Plant Battery Charger. The Generator Plant 125 Volt DC Bus normally supplies power

> to a Generator Plant 125 Volt DC distribution panel which supplies power to instrumentation and control circuits connected to 120 Volt AC Non-Interruptible Bus IC. The static switch of the Static Inverter IC is capable of supplying power from Static Inverter IC or its reserve 120 volt AC power source. In case of Static Inverter IC trouble, the static switch will automatically transfer from the output of Static Inverter IC to the reserve transformer which is supplied from Turbine Building Motor control Center lA.

The Reactor Plant 125 Volt DC Bus is capable of providing power to the Generator Plant 125 Volt DC Bus or the Diesel Building 125 Volt DC Bus through tie breakers.

4-21

i

-i ELECTRICAL POWER SYSTEMS BASES - (Cont'd)

The Diesel Building 480 Volt Essential Switchgear Bus IB supplies Reactor i

Building Motor Control Center IA and Diesel Building 480 Volt Motor Control Center. Diesel Generator IB is the emergency power supply for the Diesel [

Building 480 Volt Essential Switchgear Bus 1B. The Diesel Building 480 Volt Motor Control Center supplies power for the Diesel Building Battery Charger.

The Diesel Building Battery Charger carries the load of and charges the batteries on the Diesel Ruilding 125 Volt DC Bus.

The Diesel Building 125 Volt DC Bus is the normal power supply for Static Inverter 1B which supplies power to instrumentation and control circuits. ;

connected to 120 Volt AC Non-Interruptible Bus 1B. In case of Static Inverter IB trouble, its static switch can automatically transfer from the output of Static Inverter IB to the output of its alternate supply which is a transformer supplied by the Diesel Building 480 Volt Motor Control l Center. The Turbine Building 120 Volt Regulated Bus is the reserve feed for i the 120 Volt AC Non-Interruptible Bus 1B. -

i r

i i

I i

I I

i I

4-22 1

I

- 4/5.4= FIRE PROTECTION FIRE-DETECTION INSTRtNENTATION LIMITING CONDITION FOR OPERATION

=================================================================

As a minimum, the fire detection instrumentation for each fire

~

4.4.1 detection zone shown in Table 4.4.1 shall be OPERABLE.

APPLICABILITY: Whenever equipment in that fire detection zone ~is required-to be OPERABLE.

ACTION:

With one or more of the fire detection instruments shown in Table 4.4.1 inoperable:

a. Within I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instruments:

1) at least once per four hours in the Reactor Containment Building

~

and High Radiation areas;

2) at least once per hour in any other area,

b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or,'in lieu of a Licensee Event Report, prepare and submit a Special Report to the Commission pursuant to Specification 6.8.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERABLE status.

SURVEILLANCE REQUIREMENTS

=====u===========================================================

5.4.1.1 Each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST.

l 5.4.1.2 The supervised circuits associated with the detector alarms of each

! of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months.

i i

l l

i l

l 4-23 l

l l

t

4

/

i flRE PROTECTION

. FIRE DETECTION INSTRIMENTATION u

TABLE 4.4.1

b. FIRE DETECTION INSTRlNENTS MINIMLN-INSTRINENT LOCATION -INSTRlNENTS OPERABLE

1. Containment Building

Zone 19 Dome Level 2

= Zone 19 Mezzanine Level 2 Zone 19 Grade Level 2

, Zone 19 Basement Level 2

2. Electrical Equipment Room 2

3. Crib House, Zone 4- 1

4. Diesel Generator Rooms

-1A Zone 17 1 IB Zone 16 1-

5. ~ 1B Emergency Diesel Generator Building Battery Room Zone 16 1

6. Oil Storage Roos; Zone 13 1

7. Control Room and Office Area Zone 8 2 4-24

r FIRE PROTECTION FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION

=======================================================e=========

~4.4.2 The Fire Suppression Water System shall be OPERABLE with:

a. At least one high pressure diesel driven pump with a capacity of 750 ,

gallons per minute with its discharge aligned to the fire suppression header.

b. An OPERABIE flow path capable of taking suction from the Mississippi River and transferring the water through distribution piping with

) OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves and the first velve upstream of the water flow alarm device on each sprinkler, hose standpipe or spray system riser required to be OPERABLE per Specification 4.4.3.

APPLICABILITY: At all times.

ACTION:

With the Fire Suppression Water System inoperable:

1. Establish a backup fire suppression water system within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2. Submit a Special Report in accordance with Specification 6.8.2 within 30 days outlining the action taken, the cause of inoperability and the plans and schedule for restoring the system to OPERABLE status.

SURVEILLANCE REQUIREMEhTS

======================c===========,==============================

5.4.2.1 The Fire Suppression Water System shall be deuonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve, manual or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position,

b. At least once per 6 months by performance of a system flush,

c. At least once per 18 months by cycling each manual valve in the flow path through at least one complete cycle of full travel.

d. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

1. Verifying thet the pump develops at least 750 gpm at a system head of 320 feet, and 4-25

1

.p FIRE PROTECTION FIRE SUPPHESSION WATER SYSTEM SURVEIM ANCE - (Cont'd)

12. Verifying that the high pressure pump starts to maintain the Fire Suppression Water System pressure 1 60 psig,

e. At least once per.3 years by performing a flow test of the system in accordance with Chapter 5 Section 11, of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.

5.4.2.2 Each fire pump diesel engine shall be demonstrated OPERABLE: i

a. At least once per 31 days by verifying that the fuel storage tank for A unit contains at least 270 gallons of fuel or for B unit contains at least 108 gallons of fuel.

b. At least once per 31 days by starting the diesel from ambient conditions and operating for at least 20 minutes.

c. At least once per 92 days by verifying that a sample of diesel fuel from within three inches of the bottom of each fuel storage tank is within acceptable limits specified in Table 1 of ASTM D975-81 for water and sediment, and or viscosity,

d. At least once per 18 months, by:

1. Subjecting the diesel to the annual inspection in accordance with its manufacturer's recommendations for the class of service, and

2. Verifying the diesel starts from ambient conditions on the auto-start signal and operates for 1 20 minutes while loaded with the fire pump.

4-26

w FIRE PROTECTION FIRE SUPPRESSION WATER SYSTEM

-SURVEILLANCE REQUIREMENTS - (Cont'd) 5.4.2.3 Each fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1. The electrolyte level of each battery is above the plates, and

2. The overall battery voltage is 1 24 volta,

b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.

c. At least once per 18 months by verifying that:

1. The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and

2. The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

4-27

l H RE. PROTECTION BfBAY AND/OR SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION

======================================================s==========

4.4.3 h .following spray and/or sprinkler systems shall be OPERABLE:

a. The turbine lube oil reservoir.

b. h turbine lube oil storage room.

c. N reserve auxiliary transfoneer.

.d. lA Diesel Generator room.

e. Electrical penetration area.

APPLICABILITY: At all times.

ACTION:

With any of the above required spray and/or sprinkler systems inoperable, establish a 1-hour fire watch within I hour; restore the system to OPERABLE status within 14 days or, in lieu of a Licensee Ever.t Report, prepare and submit a Special Report to the Comunission pursuant to Specification 6.8.2 within the.next 30 days outlining the action taken, the cause of the inoperability and plans and schedule for restoring the system to OPERABLE status.

SURVEILLANCE REQUIREMENTS

=================================================================

5.4.3.1 % above required spray and/or sprinkler system shall be demonstrated OPERABLE:

a. At least once per 18 months by:

1. Cycling each manual valve in the flow path through at leest one complete cycle of full travel.

2. Perfonsing a system functional test which includes simulated automatic actuation of the reserve transformer deluge system.

3. Inspecting the spray headers to verify their integrity.

4. Inspecting each nozzle to verify no blockage.

4-?8

EftE PROTECTION CHJMICAL EXTINGUISilING LIMITING CONDITIONS FOR OPERATION

=================================================================

4.4.4 The Chemical Extinguishing System located in the IB Emergency Diesel Generator Room shall be OPERABLE with a minimum of 4 high pressure CO2 bottles weighing at least 190. pounds each.

APPLICABILITY: At all-times.

ACTION:

With the above required Chemical Extinguishing System inoperable, establish a 1-hour fire watch within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; restore the system to OPERABLE status within 14 days or, in lieu of a Licensee Event Report, prepare and submit a Special Report to the commission pursuant to Specification 6.8.2 within the next 30 days' outlining the action taken, the cause of the inoperability and the plans -

and schedule for restaring the system to OPERABLE status.

SURVEILLANCE REQUIREMENTS-

=================================================================

5.4.4 The above required Chemical Extinguishing System shall be demonstrated OPERABLE:

a. At least once per 6 months by verifying tanx weight of each of the required bottles.

b. At least once per 18 months by:

1. Verifying the system, including associated ventilation dampers, actuates manually and automatically with the solenoid valve ,

removed from the respective bottles, upon receipt of a simulated test signal, and

2. Performance of a flow test through headers and nozzles to assure l no blockage.

l 4-29 i

b@

n FIRR PROTECTION FIRE Il0SE STATIONS

,3 LIMITING CONDITIONS FOR OPERATION

================================s=========u=================================.

4.4.5 The fire hose stations in the following locations shall be OPERABLE:

1. 1B Emergency Diesel Generator Building.

2. Turbine Building Grade Floor by turbine lube oil reservoir.

3. Main Floor Turbine Building outside Control Room.

4. Containment Buililing Basement.

5. Waste Treatment Building.

APPLICADILITY: At all times.

ACTION:

With a hose station inoperable, establish a 1-hour fire watch, or route an additional hose of equivalent capacit'y to the unprotected area within one hour.

SURVEILLANCE REQUIREMENTS

=================================================================

5.4.5 Each of the fire hose stations shall be demonstrated OPERABLE:

a. At least once per 31 days by visual inspection of the station to assure all required equipment is at the station.

b. At least once per 18 months by:

1. Removing the hose for inspection and reracking, and

2. Replacement of all degraded gaskets in couplings. ;

c. At least once per 3 years by:

1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blo<kage. .

2. Conducting a hose hydrostatic test at a pressure at least 50 psig greater than the maximum pressure available at that hose station.

4-30 7

FIRE PROTECTION YARD FIRE HYDRANTS AND OUTSTDE 110SR___110USES LIMITING CONDITION FOR OPERATION

===============5=================e===============================

4.4.6 The five yard fire hydrants and four outside hose houses shall be OPERABLE.

,APPLICABIL11Y: At all times.

ACTION:

With one or more of the yard fire hydrants or outside hose houses inoperable, verify sufficient lengths of 2-1/2 inch diameter hose are available to provide service to the unprotected area within one hour. Restore the inoperable hydrant (s) and/or hose houses (s) to OPERABLE status within 14 days or, in lieu of a Licensee Event Report, prepare and submit a Special Report to the Commission pursuant to Specification 6.8.2 within the next 30 days outlining the action taken. the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

SURVEILLANCE REQUIREMENTS

===== ================================c=============:============

5.4.6 Each of the yard fire hydrant 9 and outside hose houses shall be demonstrated OPERABLE:

a. At least once per 31 days by visual inspection of the outside hose house to assure all required equipment is at the hose house,

b. At least once per 6 months, by visually inspecting each yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged,

c. At least once per 12 months by:

1. Conducting a hose hydrostatie test at a pressure at least 50 psig greater than the maximum pressure available at any yard fire hydrant.

2. Replacement of all degraded gaskets in couplings.

l 4-31 I

BASES

====a============================================================

4/5.4 FIRE PROTECTION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early

. stages. Prompt detection of fires will reduce the potential for damage to equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is I inoperable, the establishment of fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where equipment needed for safe storage is located. The fire suppression system consists of the water system, spray and/or sprinklers, 002, and fire hose stations. The collective capability of the fire suppression systems is adequate to ainimize potential damage to j l

l equipment and is a major element in the facility fire protection program. l In the event that portions of the fire suppression systems are inoperable, l alternate backup fire fighting equipment is required to be made available in I the affected areas until the inoperable equipment is restored to service. !

l In the event the Fire Suppression Water System becomes inoperable, immediate corrective measures must be taken since this system provides the major fire !

suppression capability of the plant. The requirement for a report to the l Commission provides for evaluation of the acceptability of the corrective measures to provide adequate fire suppression capability for the continued I protection of the nuclear facility.

4-32

4/5.5 INSERVICE INSPECTION 4.5.1 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION

=================================================================

4.5.1.1 The structural integrity of ASME Code Class 3 components associated with fuel storage shall be maintained in accordance with Specification 5.5.1.1.

4.5.1.2 The structural integrity of the LACBWR ventilation stack and the smoke stack of the conventional steam power generating station, Genoa 3, adjacent to the LACBWR facility, shall be maintained in accordance with Specification 5.5.1.2.

APPLICABILITY: At all times.

ACTION:

a. With the structural integrity of any ASME Code Class 3 component (s) associated with fuel storage not conforming to the above requirement, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service,

b. With the structural integrity of the LACBWR or Genou 3 stack not maintained, perform an engineering evaluation.

SURVEILLANCE REQUIREMENTS

=================================================================

5.5.1.1 Inservice inspection of ASME Code Class 3 components associated with fuel storage shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. 1974 Edition, Summer 1975 Addenda. Nothing in the ASME Boiler and Pressure Vessel Code shall be const rued to supersede the requirements of any Technical Specification.

5.5.1.2 The exterior surfaces of the Lt.CBWR ventilation stack and the Genoa 3 stack shall be inspected for structural integrity at least once every 5 l years.

l 4-33

INSERVICE INSPECTION PUMPS AND VALVES LIMITING CONDITION FOR OPERATION

=================================================================

4.5.2 The OPERABILITY of ASME Code Class 3 pumps and valves associated with

-fuel storage shall be maintained in accordance with section 5.5.2.

APPLICABILITY: At all times.

ACTION:

With a pump or valve falling into its Action range, declare the affected pump or valve INOPERABLE.

SURVEILLANCE REQUIRIJfENTS

=================================================================

5.5.2 Inservice testing of ASME Code Class 3 pumps and valves associated with fuel storage shall be performed in accordance with Section XI of the l ASME Boiler and Pressure Vessel Code,1974 Edition, Summer 1975 Addenda. !

Nothing in the ASME Boiler Pressure Vessel Code shall be construed to 1 supersede the requirements of any Technical Specification.

l l

4-34

BASES

=================================================================

4/5.5 INSERVICE INSPECTION 1his sectica ensures that inservice inspection of ASME Code Class 3 components associated with fuel storage and inservice testing of ASME Code Class 3 pumps and valves associated with fuel storage will be performed in accordance with section XI of the ASME Boiler and Pressure Vessel Code. The edition specified is that used at the time of plant shutdown.

Undet the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vesse] Code and applicable Addenda. For example, the requirements of Specification 3.7 to perform surveillance activities prior to entry into specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision whic't allows pumps to be tested up to one week after return to normal operation. And, for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Provision which allows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before being declared inoperable.

This section also ensures that the structural integrity of the LACBWH and Genoa 3 stacks are maintained.

4-35

4/5.G RADIATION MONITORING BADIATION HONITORS.

LIMITING CONDITION FOR OPERATION

=================================================================

4.6.1 Area radiation monitors and portable radiation detectors shall be demonstrated OPERABLE in accordance with Specification 5.6.1.1 for area radiation monitors and Specification 5.6.1.2 for portable radiation detectors.

APPLICABILITY: At all times.

ACTION:

With a monitor / detector not satisfactorily meeting its surveillance criteria or if its surveillance requirement is not performed, declare the monitor / detector INOPERABLE.

SURVEILLANCE REQUIREMENTS

==============e===================================================

S.6.1.1 Each area radiation monitor shall be demonstrated OPERABLE by performance of a CHANNEL CHECK at least daily, a CHANNEL FUNCTIONAL TEST at least quarterly, and a CHANNEL CALIBRATION at least once per 18 months.

5.6.1.2 Each portable radiation detector shall be dewonstrated OPERABLE by performance of a SOURCE CHECK at least once every 2 weeks and calibrated at least semiannually.

4-36

RADIATION MONITORING POST-ACCIDENT RADIATION MONITORING INSTRlMENTATION LIMITING CONDITION FOR OPERATION

=================================================================

4.6.2 The following post-accident radiation monitoring instrumentation channels shnll be~ 0PERABLE with their alarm setpoints within the specified limits.

a. A liigh Range Containment Building Area Radiation Monitor, with its high alarm setpoint at 1 100 R/hr and

b. A Stack Midrange Noble Gas Effluent Monitor with its high alarm setpoint at 1 1.0 E-1 pci/cc or equivalent counting rate to Kr-85.

APPLICADILITY: At all times.

ACTION:

a. With a p >st-accident radiation monitoring channel alarm setpoint exceeding the above value, adjust ihe setpoint to within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the channel INOPERABLE.

b. With less than the required post-accident radiation monitoring instrumentation channels OPERABLE, restore the INOPERABLE channel (s) to OPERABLE status within 7 days.

SURVEILLANCE REQUIRDIEhTS

==e===================================2==========================

5.6.2 Each post-accident radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of a CIIANNEL CIIECK daily CllANNEL PUNCTIONAL TEST monthly, and CilANNEL CALIBRATION at least once per 18 months.

l l

4-37

RADIATION MONITORING SEALED SOURCE CONTA'jjNATION LIMITING CONDITION FOR OPERATION

====n============================================================

-4.6.3 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting materials or 5 microcuries of alpha enitting material shall be free of 1 0.005 microcuries of removable contamination.

APPLICABILITY: At all times.

ACTION:

n. Each sealed source with removable contamination in excess of the above limit shall be immediately withdrawn from use and:

1. Either decontaminuted and repaired, or

2. Disposed of in accordance with Commission Regulations.

SURVEILLANCE REQUIREMENTS

====,============================================================

5.6.3.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a. The licensee, or

b. Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

5.6.3.2 Test Frequencies - Each category of sealed sources, excluding i startup sources, fission detectors previously subjected to core flux, and sources as specfied in 10CFR30.15 shall be tested at the frequency described below,

a. Sources in Use - At least once per six months for all scaled sources '

containing radioactive materials; ,

1. With a half-life greater than 30 days, excluding !!ydrogen 3, and

2. In any form other than gas.

b. Stored Sources Not in Use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources transferred without a certificate indicating the last test date shall be tested ;

prior to being placed into use.

i 4-38 i

, _; ' T d I

RADIATION MONITORING ;

{LA1.BD SOURCE CONTAMINAllQM SURYRILLANCE - (Cont'd) 5.6.3.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source leakage tests reveal the presence of

~

1 0.005 microcuries of removable contamination.

l i

L t

i k

t I

I 1 s L

I t

k I

i l

I 4-39 L

4/5.0 RADIATION MONITORING BASES m:23:z2suse:r=====22s== mex ::::====n=====e=====a=========t=======::::::=====

4/5.6.1 Radiation _ Monitors The surveillance requirements specified will ensure that area radiation monitors and portable radiation detectors being used are functional.

4/5.6.2 Post-Accident Radiation Monitoring Instrumentation Thi operability of the post-accident radiation monitoring instrumentation is established so that if an accident does occur, it can be monitored. The setpoints are based on analyses of the radiological aspects of selected worst case accident scenarios involving the release of Kr-85 from spent fuel assemblies and the direct radiation from exposed irradiated components.

The surveillance requirements specified for these systems ensure that the everall system functional capability is maintained comparable to the original design standards. The periodic surveillance testa performed at the minimum frequencies are sufficier.t to demonstrate this capability. Calibrations of Containment Building Aren High Range Radiation Monitors way be performed under the requirements of NUREG-0578.

4/S.6.3 Scaled Source Cont aminntion The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for

, plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

Sealed sources are classified into two groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

4-40 l

I

4/5.7 RADIOACTIVE EFFLUERTS 4/5.7.1 BADI0 ACTIVE LIQUID EFFLUENTS INSTRUMENTATION LIMITING CONDITION FOR OPERATION x======================e=========e===========e=====.,========================

4.7.1.1 The following radioactive liquid effluent monitoring instrumentation channels shall be OPERABLE, with their alarm setpoints set to ensure that the limits of Specification 4.7.1.2 are not exceeded,

n. Liquid Radwaste Effluent Line Monitor or Turbine Condenser Cooling Water Monitor, and

b. Liquid Radwaste Effluent Line Flow Meter.

The alarm setroints for these monitors will be determined and adjusted using methodology in the ODCM.

APPLICABILITY: At all times when releasing liquid radioactive effluents.

ACTION:

n. With the Liquid Radwaste Effluent Line Monitor or Turbine Condenser cooling Water Monitor channel alarm /trippoint setpoint less conservative I than that required by the above spe cification, insediately suspend the release or declare the channel inoperable or change the setpoint so that it is acceptably conservative.

b. With both channels not OPERABLE, or if both alarm setpoints are found to be less conservative than required, suspend release of liquid radioactive effluent without delay. Effluent releases may be resumed with neither activity monitor OPERABLE. provided that at least two independent samples are analyzed and that at least two technically qualified members of the staff independently verify the release rate calculations. If channels are not operable for more than 30 continuous days, explain in the next Semi-Annual Effluent Reports pursuant to Specificat ion 6.8.1.2.

c. With the flow meter not OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves may be used to estimate flow.

SURVEILLANCE REQUIREMENTS l ============================================================s================

5.3.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by perfonnance of the CHANNEL CHECK, SOURCE CHECK, Ct!ANNEL FUNCTIONAL TEST, and CilANNEL CALIBRATION operations at the j frequencies abown on Table 5.7.1.1.

l l

l 4-41 i

r,M

+

TABLE 5.7.1.1 EADIOACTIVE LIQUID EFFLUENT MONITORINT INSTEDENTATION SURVEILLANCE REQUIEDENTS

, CHANNEL SURVEILLANCE' g CHANNEL- SOURCE FUNCTIONAL- CHANNEL REQUIRDECNT

- INSTRt#ENT CHECK CHECK. TEST CALIBRATICW COISITIONS

c. Liquid Radwaste Effluent Line Monitor P P* . 00 ) ~ R(3) **

b. Turbine Condenser Cooling Water Monitor P M* ' 00') - R(8) **

c. Liquid Radweste Effluent Line Flow Meter D(*) NA NA R(4) **

~

Background radiation may be used for the source check.

c* During applicable conditions.

(I) he CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any Lof the '

following conditions exist

a. Instrument indicatu measured levels at the alarm setpoint. .

b. Instrument indicates a downscale (circuit failure) failure.

(2) CHANNEL CHECK shell consist of verifying indication of flow during periods' of release. CHAlWEL CIRCK shall be ==rta at lear.t once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or days in which continuous, periodic, or batch releases are made.

(3) The CHANNEL CALIBRATION shall include the use of a known liquid radioactive source positioned in a reproducible ,

geometry with respect to the sensor. The source will have the g = = emitting radionuclide mixture'and'activityi concentration which would normally be measured by the. channel during batch discharges. ,

(4) The CHANNEL CALIBRATION will be in accordance with plant procedure.

l i

i 442

-,_~-__._._.._.___...__,,.__.._..,1., _ _ - - . . - - . -- _ , _ . . _ _. .- . .u __ ....= __

. }tADI0 ACTIVE EFFLUENTS

}[AD10 ACTIVE JIQUID EFFLUENTS

-CONCENTRATION LIMITING CONDITION FOR OPERATION ar=======us===x ==amev=====================ss================23 eenussanneere 4.7.1.2 The concentration of radioactive material released in liquid effluents at any time to areas beyond EFFLUENT RELEASE BOUNDARY shall be limited to the concentrations specified in 10 CFR Part 20 Appendix B.

Table II. Column 2, i'. sue of December 30, 1982, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to G x 10-4 pei/ml total activity concentration.

APPLICABILIIY: At all times.

AC TION:

With the concentration of radioactive material released beyond the EFFLUENT RELEASE BOUNDARY exceeding the above limits, without delay restore concentration to within the above limits.

SURVEILLANCE REQUIREMENTS

===ers== == a22:e======== r===nusmermaw=r=== scrars=====neestar==s=====

5.7.1.2 The radioactivity content of each batch of radioactive liquid waste to be discharged shall be determined prior to release by sampling and analysis in accordance with Table 5.7.1.2. The results of pre-release analyses shall be used in accordance with the ODCM methodology to assure that the concentration at the point of release is maintained within the limits of Specification 4.7.1.2.

RADIOACTIVE LTQUID EFFI,UENTS DOSE LIMITING CONDITION IVR OPERATION

================ =======z=============================2==========

4.7.1.3 The dose or dose comitment to a MF}mER OF THE PUBLIC from radioactive materials in liquid effluents released to areas beyond EFFLUENT RELEASE BOUNDARY shall be limited:

a. During any calendar quarter to 1 1.5 mreas to the total body and to i

1 5 areas to any organ, and l b. During any calendar year to I 3 mress to the total body and to i 10 mress to any organ.

l 4-43 i

19ADIOACTIVR ELFLUENTS EADIOACTIVR LIQUID EFFl.l%NTS - (Cont'd)

APPLICABILITY At all times.

49Il0Nt With the calculated dme from the release of radioactive _ materials in liquid effluents exceeding the above-limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report which identifies the cause(s) for exceedint - 'imit(s) and defines the corrective ,

actions which have been or will be tt assure that subsequent releases shall be in compliance with the abov: 4.

-5.7.1.3 Dose _Calcu19tions: Cumulati. dose contributions from liquid effluents shall be determined at least once per calendar quarter in accordance with the ODCM.

f O

l l

l T

1 k

4-44 i l - _

g,~~ _

1 g,j ~"

.1 9 ;. ,

11 ' RADIOACTIVE EFFLUENTS' LRADI0 ACTIVE LIQUID EFFLUENTS F : TABLE 5.7.1.2-o .. .

"RADIOACTIVE LIQUID WASTE SAMPLING AND' ANALYSIS PROGRAM l LIQUID- :- l MINIMUM- :. TYPE OF :

RELEASE- SAMPLING -l ANALYSIS- ! ACTIVITY :

iTYPE : ~- FREQUENCY : FREQUENCY- . l-ANALYSIS (d) l l l ! : :

!. Waste Tank l P =: P Principal :

Batch '

l Gammu :

! Releases (a) : : : Emitters (c)

One Batch /M :M l Dissolved and 11 : : : Entrained :

l l l l Gases

l l (gamma '

: : emitte -f

P : .M l r.-3 :

: ! Composite (b) ,__ Gross Alphal l !P : Q S r-90 :

: : Composite (d) : ;

l l Fe-55 :

l l l l l

a. A batch release is the discharge of liquid wastes of a discrete volume.

Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, to assure representative sampling.

b. A composite = sample is one made up of individual samples which are 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 liquid release,

c. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Co-60, Zn-65, Cs-134, Cs-137, and Ce-144. 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.

'd.. Methods of calculating the Lower Limits of Detection (LLD) shall be contained in plant. procedures and are calculated in accordance with criteria of NUREG-0473, Revision 2.

f 4-45

~. _ - - - -

RADIOACTIVE EFFLUENTS 4/5.7.2 RADIOACTIVE GASEOUS EFFLUENTS INSTRUMENTATION LIMITING CONDITION FOR OPERATION

===================================================================== -

4.7.2.1 The radioactive gaseous effluent monitoring instrumentation .nannels shown in Table 4.7.2.1 shall be OPERABLE with their alarm rmd/or trip setpoints set to ensure that the limits of Specification 4.7.2.2 are not exceeded. The stack noble gas instrumentation alarm setpoint will be determined and adjusteo in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: As shown in Table 4.7.2.1 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm and/or trip setpoint less conservative than that required by the above specification, declare the channel inoperable or change the setpoint so that it is acceptably conservative.

b. With less than the minimum number of radioactive gascous effluent l monitoring instrumentation channels OPERABLE, take the ACTION required

) by Table 4.7.2.1. Exert best efforts to return the instruments to l OPERABLE status within 30 days, and if unsuccessful, explain in the next Semi-Annual Radioactive Effluent Release Report pursuant to 6.8.1.2 why the inoperability was not corrected in a timely manner.

SURVEILLANCE REQUIREMENTS

=================================================================

5.7.2.1 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST, and CRANNEL CALIBRATION operations at the frequencies shown in Table 5.7.2.1.

l 4-46

7, . -- -.,v --. -

RADIOACTIVE EFFLUENTS!

.s

RADIOACTIVE GASEOUS EFFLUENTS L

t-

, TABLE 4.7.2.l' r

, RADIOACTIVE'GASE0US' EFFLUENT MONITORING INSTRUMENTATIONS MINIMUM CHANNELS APPLICABLE INSTRUMENT OPERABLE CONDITIONS ACTION

- 1. - Reactor Containment Building Ventilation Monitor System za. Particulate Activity 1

B Monit'or b .~ Gaseous Activity Monitor- 1

B

c. Sampler Flow Rate 1

C Measuring Device l2. Stack Monitor System

a. -Noble Gas Activity Monitor 1 **- D

-b. Particulate Activity Monitor 1 ** E

c. Sampler Flow Rate 1 ** C

- Measuring Device

-When Containment Building Ventilation System is in operation.

- -At all times, unless alternate monitoring is available.

A. .For post-accident instrumentaion, refer to Specification 4.6.2.

c . B. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases through this pathway may continue as long as stack monitors are OPERABLE and CONTAINMENT

. INTEGRITY is not required; otherwise, isolate Containment Building Ventilation.

C. 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 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D. With the-number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the Containment Building Monitor Gaseous Activity Monitor is OPERABLE; otherwise, isolate the contninment Building.

E. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided continuous collection of samples with auxiliary sampling equipment is initiated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4-47

1 TABLE 5.7.2.1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL SURVEILLANCE CHANNEL SOURCE FUNCTIONAL CHANNEL <4) REQUIREMENT INSTRUMENT CHECK CHECK TEST CALIBRATION CONDITIONS

1. Reactor Containment Building Ventilation Monitor System

a. Particulate Activity Monitor D M Q(1) R *

b. Gaseous Activity ffonitor D M Q(1) R *

c. Sampler Flow Hate Measuring Device D N/A Q(3) R *-

2. Stack Monitor System

a. Noble Gas Activity Monitor D M Q(2) R *-

b. Particulate Activity Monitor D N/A Q<2) R *

c. Sampler Flow Hate Measuring Device D N/A Q(3) R *

During applicable conditions per Table 4.7.2.1.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway (if CONTAINMENT ,

INTEGRITY is required) and control room alarm annunciation occurs if any of the following conditions ~ exist:

a. Instrument indicates measured levels at or above the alarm setpoint.

b. Instrument indicates a downscale failure (provides control room annunciation alarm only).

c. Instrument indicates a circuit failure (provides control room annuniciation' alarm only).

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of' the following conditions exist:

a. Instrument indicates measured level above the alarm setpoint on one channel,

b. Instrument indicates a failure by a Low Flow and Low Count Rate signal.

c. Instrument controls in Maintenance mode.

(3) The CHANNEL FUNCTIONAL TEST shall also demonstrate that the control room local alarm occurs if the flow instrument indicates measured levels below the minimum and/or above the maximum alarm setpoint.

(4) The CHANNEL CALIBRATION shall be conducted in accordance with plant procedures.

4-48 m

r- _

' RADIOACTIVE EFFLUENTS -

, RADIOACTIVE GASEOUS EFFLUENTS

-INSTANTANEOUS DOSE RATE

LIMITING CONDITION FOR OPERATION ,

.====================================s========================================

4.7.2.2 ' The dose rate due to radioactive materials released in gaseous stack effluents to areas beyond the EFFLUENT RELEASE BOUNDARY shall be limited to the . following:

a. The dose . rate limit for noble gases shall be 1500 mrems/ year to the total body and 1 3000 mrems/ year to the skin, and

b. The dose rate limit for H-3 and for all radionuclides in particulate form with half lives greater than 8 days, shall be I 1500 arems/ year to any organ.

-APPLICABILITY: At all' times.

ACTION:

With the dose rate (s) exceeding the above limits, without delay decrease the release rate to within the above limit (s).

' SURVEILLANCE REQUIREMENTS

=================================================================

5.7.2.2.1 The dose rate due to noble gases in gaseous stack effluents shall be determined to be within the above limits in accordance with the ODCM.

5.7.2.2.2 The dose rate due to H-3 and for all radioactive materials in particulate form with half lives > 8 days in gaseous stack effluents shall be determined to be within the above limits in accordance with the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 5.7.2.2.

4-49

e.

/-

V TABLE 5.7.2.2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM

Gaseous : Sampling : .: _

1:

Release Type : Frequency ~ : Minimum Analysis Frequsney : Type of Activity Analysis (d)(*)

: W(*) Composite ;

A. Stack : Continuous <b) : Particulate Sample : Principal Gamma Emitters (c) -

Effluents : Q Composite : :: ,

l Continuous <b) : Particulate Sample Sr-90 ':

: : Q Composite :

: Continuous Particulate Sample ~ ! Gross Alpha : <

: Noble Gas : Noble Gases :-
Continuous <b) : Monitor : Gross Beta and Gamma- .
:M : ': :
: Grab Sample :M : H-3(d) :
TABLE NOTATION:
(a) Filter samples shall be changed at least. weekly, and filter analyses shall be completed within 7 days.
(b) 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 5.7.2.2.1, 5.7.2.2.2, 5.7.2.3 and 5.7.2.4. . .
(c) The principal gaimea emitters for which the LLD specification applies exclusively are the following radionuclides:
Mn-54, co-60, 2n-65, Cs-134, Cs-137, and Ce-144 for particulate emissions. This list does not mean that only these.
nuclides are to be considered. Other gammaa peaks that are identifiable' and measurable, together with those of the .
above nuclides, shall also be analyzed and reported in.the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.8.1.2.
~
(d) When upper cavity is flooded or FUEL HANDLING is being performed, stack tritium grab samples lwill be'taken~at least once per 7 days.
(e) Lower Limits of Detection (LLD) are determined in accordance with plant procedures and are calculated in accordance with criteria of NUREG-0473, Revision 2.
t 4 -
?? ,
,; -RADIOACTIVE EFFLUENTS.
RADIO 5CTIVE GASEOUS EFFLUENTS DOSE. NOBLE GASES LIMITING CONDITION FOR OPERATION
=================================================================
-4.7.2.3 The air dose.to a MEMBER OF THE PUBLIC due to~ noble gases released U in . gaseous effluents to' areas beyond EFFLUENT RELEASE BOUNDARY shell be limited to the following, -(See ' Figure 4/5.7):
a. During any calendar quarter, to 1 5 mrads-for gamma radiation and 110 mrads for beta particle radiation; and
b. During any calendar year, to 1 10 mrads for gamma radiation and 120 mrads for beta particle radiation.
APPLICABILITY: At all times.
-ACTION:
With the calculated air. dose from radioactive nobic gases in gaseous effluents exceeding any of the above limits, prepare and submit to the -
Commission within 30 days, pursuant to specification 6.8.2, a Srecial Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions which have been taken or will be taken to reduce the-releases of radioactive noble gases in gaseous effluents so that the cumulative dose during each subsequent-quarter is within 5 mrads for gamma
. radiation and 10 mrads for beta radiation, and during the calendar year is
- within 10 mrads for gamma radiation and 20 mrads for beta radiation.
SURVEILLANCE REQUIREMENTS
================================================================
5.7.2.3 Dose Calculations: Cumulative dose contributions shall be determined in accordance with the ODCM at least quarterly.
4-51
s l RADIOACTIVE EFFLUENTS
' RADIOACTIVE GASEOUS EFFLUENTS
' DOSE RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION
=================================================================
4.7.2.4 The_ dose to a MEMBER OF THE FUBLIC from H-3, and all radionuclides in particulate form with half-lives greater than 8 days, in gaseous effloents released to areas beyond EFFLUENT RELEASE BOUNDARY rhall be limited to the following (See Figure 4/5.?):
a. During any calendar quarter to I 7.5 mrems to any organ, and
b. _During any calendar year to 1 15 mrens to any-organ.
APFLTCABILITY: At all times.
ACTION:
With the calculated dose from the release of 11-3 and all radionuclides in
-particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30' days, pursuant to Specification 6.8.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions which have been taken or will be taken to reduce these releases in gaseous effluents during remaining quarters so that the cumulative dose during each subsequent quarter is within 7.5 mrems and for the calendar year is within 15 mrems to any organ.
SURVEILLANCE REQUIREMENTS
=================================================================
5.7.2.4 Dose Calculatipns: Cumulative dose contributions shall be determined in accordance with the ODCM at least quarterly.
4-52
7.- ,
- At ,
RADIOACTIVE EFFLUENTS-SOLID RADIOACTIVE WASTE
. LIMITING CONDITION FOR OPERATION-
= ===============================================================
~4.7.3 Solid' radioactive wastes shall be handled in accordance with a PROCESS CONTROL PROGRAM in order to meet shipping and burial ground = requirements.
APPLICABILITY: At all times when. processing solid radioactive wastes for shipment aled disposal.
- ACTION:
With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of det'ectively processed or defectively packaged solid radioactive wastes from the site.
SURVEILLANCE REQUIREMENTS
=================================================================
5.7.3 The PROCESS CONTROL PROGRAM shall be used to assure the appropriate form for packaging each type of radioactive waste (e.g., filter sludges, spent resins, tank bottoms, dry active wastes).
4-53 A* .C t a* -
RADIOACTIVE EFFLUENTS
. TOTAL DOSE LIMITING CONDITION FOR OPERATION
=================================================================
4.7.4 The dose equivalent to any MEMBER OF THE PUBLIC due to releases of radioactivity and radiation, shall be limited to 1 25 mrems to the total body or any organ (except the thyroid, which is limited to I 75 mrems) over a period of one calendar year.
APPLICABILITY: At all times.
ACTION:
With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limit of Specifications 4.7.1.3.b.
4.7.2.3.b or 4.7.2.4.b, a determination should be made, ir.cluding direct radiation from reactor containment and radioactive waste storage tanks to determine if limits of Specification 4.7.4 have been exceeded. If the limits of Specification 4.7.4 have been exceeded, prepare and submit a Special Report (including an analysis which estimates the radiation exposure to a MEMBER OF THE PUBLIC for the calendar year) to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 4.7.4. If the release condition resulting in violation of Specification 4.7.4 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190.
Submittal of the Special Report is considered a timely request, and a variance is granted until staff action on the request is complete.
SURVEILLANCE REQUIREMENTS
=======================================r:===========:=============
5.7.4.1 Dose Calculations: Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 5.7.1.3, 5.7.2.3, 5.7.2.4, and in accordance with the Offsite Dose Calculation Manual (ODCM) once per year.
5.7.4.2 Dose Determination: Cumulative dose contributions from direct radiation from the reactor containment or radioactive waste storage tanks shall be determined in accordance with the methodology and parameters of the ODCM once per year.
4-54
4/5.8 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM AND INTERLABORATORY COMPARISON LIMITING CONDITION FOR OPERATION
=================================================================
4.8.1 The Radiological Environmental Monitoring Program, shall be conducted as specified in Table 4.8.1-1. An Interlaboratory Comparison Program for annual analyses of radioactive materials shall be conducted.
APPLICABILITY: At all times.
ACTION:
a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 4.8.1-1, prepare and subrit to the Commission, in the Radiological Environmental Monitoring Report pursuant to Specification G.8.1.1.c, a description of the reasons for not conducting the program as required, analysis of the causes of unexpected results, and the plans for preventing a recurrence.
b. With radiological environmental sample analyses in excess of the reporting levels in Table 4.8.1-2, when averaged over any calendar quarter, prepare and submit to the Commission a Special Report within 30 days, parsuant to Specification 6.8.2, with a description of the reasons for exceeding the reporting levels.
c. With Interlaboratory comparisons not being performed, report the corrective actions taken to prevent a recurrence to the Commission in the Radiological Environmental Monitoring Report, pursuant to Specification 6.8.1.1.c.
. RADIOLOGICAL ENVIRONMENTAL MONITORING SURVEILLANCE REQUIREhTNTS
========================================================a========
5.8.1.1 The radiological environmental monitoring samples shall be collected in accordance with Table 4.8.1-1 from locations specified in the ODCM, and shall be analyzed in accordance with requirements listed in Tables 4.8.1-1 and 4.8.1-2.
5.8.1.2 A summary of the results obtained from the above required Interlaboratory Comparison Program shall be included in the Radiological Environmental Monitoring Report pursuant to Specification 6.8.1.1.c.
4-55
%d TABLE 4.8.1-1 -
RADIOLOGICAL ENVUtDSMRSTAL MONITORING PROGRAM EXPOSURE PATHWAY TYPE AND FREQUENCY 8 AND SAMPLING NINBER OF SAMPLESW SAMPLING FREQUENCY OF ANALYSIS (1) Airborne 3 Continuous operation of sampler Analyze for gross beta activity Particles with sample collected as > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following filter determined by dustloading but at change. Gamma isotopic analysis least weekly. of each particulate sample which exhibits >10 times the mean beta activity of control station.
Gaimna isotopic analysis of at least a quarterly composite from each location.
(2) Direct Radiation 8 SA Gamma dose - at least At least 2 monitors semi-annually.
at each location.
(3) Waterborne River Water 2 M Gamma isotopic analysis on each sample monthly. Tritium analysis on composite of monthly samples at least quarterly.
(4) River Sediment 2 SA Gamma isotopic analysis of each sample.
(5) Ingestion (a) Milk 1 At least monthly when animals Gamma isotopic analysis on each are in pasture (May-Oct.). sample.
(b) Fish 1 At least semi-annually. One Gamma isotopic analysis edible sample of two different species portions of each.
in the area that are important i as a game or commercial species.
(c) Vegatation 1 At time of harvest. Gamma isotopic analyses of edible portions of ecch sample.
O Locations are specified in ODCM.
8 LLD's are calculated in accordance to criteria of, and are essentially the same as those found in NUREG-0473, Revision 2.
4-56
y .,
TABLE 4.8.1-2 REPORTING LEVELS FOR RADIOACTIVITY"CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting levels
: Water : Airborne Particulate : Fish : Milk
Analysis : pCi/l ! (pCi/m 3 ) : (pCi/Kg, Wet) ! (pCi/l)
! H-3 : 2 x 104 : : : :
Mn-54 : 1 x 103 : : 3 x 104 : :
Co-60 : 3 x 102 : : 1 x 104 : :
Zn-65 : 3 x 102 : : 2 x 104 : :
Cs-134 : 30 : 10 : 1 x 103 ': 60 :-
Cs-137 : 50 : 20 l 2 x 103 : 70 ?
4-57
.u--
y
/
4/5.7 RADIOACTIVE EFFLUENTS /
BASES
===================e==========================================n==
4/5.7.1 RADIOACTIVE LIQUID EFFLUENTS 4/5.7.1.1 INSTRUMENTATIO_N The radioactive liquid effluent instrumentation is provided to monitor the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents with the alarm setpoints set to ensure that the alarm will occur prior to exceeding the limits of 10 CFR Part 20.
4/5.7.1.2 CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. The concentration limit for noble gases is based upon the assumption that Kr-d5 is the cont rolling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
4/5.7.1.3 DOSE This specification is provided to imple* ment the requirements of Sections II.A. III. A, IV.a and Arnex of Appendix I, 10 CFR Part 50. The dose calculations in the ODCM implement the requirement in Section III.A of Appendix I 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.
4/5.7.2 RADI0 ACTIVE GASEOUS EFFLUENT!
4/5.7.2.1 INSTRINENTATION The radioactive gaseous effluent 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 only significant noble gas remaining is Kr-85. The alarm setpcints for these instruments shall be set to ensure that the alarm will accur prior to exceeding the limits of 10 CFR Part 20.
4-58
=
RADIOACTIVE EFFLUENTS BASES - (Cont'd) 4/5.7.2.2 INSTANTANEOUS DOSE RATE
.This specification is provided to ensure that the dose rate at any time at the EFFLUENT RELEASE BOUNDARY from gaseous effluents from LACBWR will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable nssurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, outside the EFFLUENT RELEASE BOUNDARY to annual avecage concentrations exceeding the limits specified in Appendix B, Table II of Id CFR Part 20 (10 CFR Part 20.106(b)). For individuals who may at times be within the EFFLUENT RELEASE BOUNDARY, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor cSove that for the EFFLUENT RELEASE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the EFFLUENT RELEASE BOUNLARY to i 500 arem/ year to the total body or to 1 3000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding organ dose rate above background to an infant via the inhalation pathway to 1 1500 mrem / year.
4/5.7.2.3 DOSE. NOBLE GASES This specification is provided to implement the requirements of Sections II.B. III. A. and IV. A of Appendix I,10 CFR Part 50. The dose calculations in the ODCM implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated.
4/5.7.2.4 LOSE. RADIONUCLIDES OTHER THAN NOBLE GASES This specification is provided to implement the requirements of Sections II.C, III.A IV.A and Annex of Appendix I, 10 CFR Part 50. The ODCM 3
calculational methods specified in the surveillance requirements implement the requirements in Section I _.a of Appendix I 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.
4/5.7.0 SOLID RADIOACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and applicable portions of 10 CFR Part 61.
4-59
g <
RADIOACTIVE ~ EFFLUENTS BASES --(Cont'd)
=================================================================
4/5.7.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 100.
]
.The specification requires the preparation and submittal of a Special Report.
whenever the calculated' doses from plant radioactive effluents exceed twice' the design objective doses of Apperdix I. The Special Report will describe a course of. action which should result in'the limitation of dose to a real individual for 12 consecutive months to within the 40 CFR-190 limits.
l l
l l
l l
4-60
/
-4/5.8 RADIOLOGICAL ENVIRONMENTAL MONIT0HING PROGRAM AND INTERLABORATORY COMPARISON BASES 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 individuals resulting.from plants effluents. This monitoring program theory supplements the' radiological effluent monitoring program by verifying that the measurable-concentrations of radioactive materials and levels of radiation are not-higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.
The requirement for participation in an Iniurlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive materini in environmental samples are performed to demonstrate that the results are reasonably valid.
4-61
Figure 1 SITE MAP INCLUDING EFFLUEhT RELEASE BOUNDARY agical' Monitoring e- s,1 l' l y b
n, ff
(
, V -
. Plant Stack (107 m) j CW Intake pe oS EXCLUSION AREA 10 M DARY ( B) 4 Y9 g ,,,3 o 1 7.
,/
/ hdi 100m -
EFFLdENT RELEASE ( =
/ -
fEOUNDARY (ERS) - -
CW Outfall ,l 1
1 a
\
y
, i o
.e.
l q
e
[ ]
j.
l Moult $
& y ,
r e gl ,
/
/
I.
f -
t i 1
Y _4 -
4-62 ODCH
Figure 1 SITE MAP INCLUDING EFFLUENT RELEASE BOUNDARY t.
/ j l l
o y l 0 " Meteorological Monitoring (
SCALC ,.
\
0 400' eoo
p ;
-l
. /
l , V -
~~
. Plant Stack (107 )
j CW Intake ' ,,g g EXCLUSION AREA / -
BOUNDARY (EAB) L6ts.s O 4 - 5
, Nh '4 s' ,
,/ , ',/ NM100m '
EFFL ENT RELEASE : i" i
\'
,/ BOUNDARY (ERB) g CW Outfall n
gn
~ '
, & l I
? W i j j' ,q \wousts 5
l l j 1
i Y el 4-62 ODCM
6. ADMINISTRATIVE CONTROLS
=================================================================
6.1 RESP 0NSIBILITY 6.1.1 The Plant Superintendent shall be responsible for overall facility-operation and shall delegate in writing the succession to this responsibility
~during his absence.
6.1.2 The Shift Supervisor (or during his absence from the Control doom, a designated individual) shall be responsible for Control Room command function.
6.2 ORGANIZATION 6.2.1 FACILITY STAFF 6.2.1.1 The facility organization shall be as follows:
a. Each on duty shift shall be composed of at least one Shift Supervisor, who is a Certified Fuel llandler, and one qualified Control Room Operator.*
b. A qualified Control Room Operator or Shift Supervisor shall be within visual and/or audio distance of the Control Room annunciators.
c. All FUE!.IIANDLING shall be directly supervised by a Certified Fuel Handler.
d. "An individual qualified in radiation protection procedures shall be onsite when there is fuel onsite or there is a potential for release of radioative materials." At least one additional Operator and one Health Physics Technician shall be onsite when spent fuel or a spent fuel shipping cask is being handled or when any evolutions are being conducted in or above the Fuel Element Storage Well,
c. A Fire Brigade of at least 3 members shall be maintained on site at all times.**
Shif t crew composition may be one less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. This provision does not permit any shift crew position to be unfilled upon shift change due to an oncoming shift crew member being late or absent.
Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of Fire Brigade members provided immediate action is taken to restore the Fire Brigade to within the minimum requirements. This provision does not permit any Fire Brigade position to be unmanned upon shift change due to an oncoming Brigade member being late or absent.
G-1
ADMINISTRATIVE CONTROLS - (Cont'd)
=================================================================
6.2.1.2 OVERTIME POLICY The working hours of Operators, the Duty Shift Supervisor, Mechanical Maintenance and Instrument & Electrical Technicians when performing duties which may affect nuclear safety, and llealth Physics Technicians, when performing radiation protection duties which may affect the safety of the public shall be limited.
In the event overtime must be used, the following restrictions shall be followed:
1) The specified personnel shall not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shif t turnover time.
2) The specified personnel shall not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-period, more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period.
3) A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months shall be allowed following overtime before the next scheduled shift for the specified personnel, if the above limits are exceeded.
In the event overtime must be used in excess of the above restrictions, the Plant Superintendent or his designate, must authorize the deviation and the cause must be documented.
L 6-2
ADMINISTRATIVE CONTROLS - (' Cont'd)-
=================================================================
6.3 -FACILITY STAFF QUALIFICATIONS _
6.3.1 Each member of-the facility staff shall meet or exceed the'ainimum
-qualifications of ANSI N18.1-1971 for_ comparable positions except for the Health Physics Supervisor. who sha11' meet or exceed the qualificatiors of Regulatory Guide 1.8, September 1975.-
6.4 TRAINING 16 . 4 . 1 A retraining and. replacement training program for the. facility staff shall be maintained which shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971. ,
6.4.2 A training program for the' Fire Brigades shall meet the requirements of Section 27 of the NFPA Code-1975, _except for the frequency of Fire Brigade ,
training sessions, which shall be held at least once per 92 days. '
6.5 REVIEW AND AUDIT 6.5.1 OPERATIONS REVIEW COMMITTEE (ORC) i FUNCTION l
6.5.1 -The Opera'.lons Review Committee shall function to advise the Plant Superintendent on all matterm related to nuclear safety.
COMPOSITION 6.5.1.2 The Operations Review Committee shall be composed of the following: I
-Chairman: -Plant Superintendent Members: -LACBWR Department Supervisors LACBWR Staff Engineers LACBWR Shift Supervisors LACBWR Management Personnel ALTERNATES' ,
l 6.5.1.3 All alternate members shall be appointed in writing by the ORP Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in ORC activities at any one time. r MEETENG VREQUENCY ,
6.5.1.4 The ORC shall meet at least once per calendar quarter and as !
convened by the ORC Chairman or his designated alternate.
ii i
L G-3 j
F -
ADMINISTRATIVE CONTROLS (Cont'd)
========================================================se=======
QUORUM 6.5.1.5 The minimum quorum of the ORC necessary for the performance of the ORC responsibility and authority provisions of the Technical Specifications shall consist of the Chairman, or his designated alternate, and 3 members, including alternates.
RESPONSIBILITY 6.5.1.6 The Operations Review Committee shall be responsible for:
a. Review of (1) all procedures required by Specification 6.6 and changes thereto, (2) any other proposed procedures or changes thereto as determined by the Plent Superintendent to affect nuclear safety.
b. Review of all proposed tests and experiments that affect nuclear safety.
c. Review of all' proposed changes to the Appendix "A" Technical Specifications.
d. Review of all proposed changes or modifications to facility systems or equipment that affect nuclear safety,
e. Investigation of all violations of the Technical Specifications including the preparation arid -forwarding of reports covering evaluation and recommendations to prevent recurrence to the General Manager and to the Safety Review Committee (SRC).
f. Review of all REPORTABLE EVENTS.
g. Review of facility operations to detect potential nuclear safety hazards.
h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant Superintendent or SRC.
i. Review of the Contingency Plan, the Plant Security Plan, and implementing procedures.
J. Review of the Emergency Plan and implementing procedures.
k. Review of proposed changes to the Process Control Program.
1. Review of the Decommissioning Plan.
6-4
ADMINISTIMTIVE CONTROLS ' Cont'd)
=======:=======================================================
AUTif0RITY 3.5.1.7 The Operatio.is Review Committee shall:
a. Recommend in writing to the Plant Superintendent approval or disapproval of items considered under 6.5.1.6a through d above,
b. Render determinations in writing with regard to whether or not rach item considereA under Specification 6.5.1.6a through e above constitutes an unreviewed safety question.
c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the General Manager and the SRC of disagreement between the ORC and the Plant Superintendent; however, the Plant Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
PECORDS 6.5.1.8 The Operations Review Committee shall maintain written minutes of each ORC meeting that, at a minimum, document the results of all ORC activities performed under the responsibility and authority provision of these Technical Specifications. Copies shall be provided to the General Manager and the SRC.
6.5.2 SAFETY REVIEW COMMITTEE (SRC)
FUNCTION 6.5.2.1 The SAFETY REVIEW COMMITTEE shall function to provide independent review and audit of designated activities in the areas of:
a. FUEL llANDLING operations,
b. Nuclear engineering,
c. Chemistry and radiochemistry,
d. Metallurgy,
e. Instrumentatioa and control,
f. Radiological safety,
g. Mechanien1 and electrien1 engineering, and
h. Quality assurance practices.
l r
6-5 ;
. .n e
ADMINISTRATIVE CONTROLS -'-(Cont'd)
.=============================================================================
cCOMPOSITION 6.5.2.2 The SRC shall be composed of the:
Chairman: Consultant Members: Consultant Plant' Superintendent Director of External Relations General Manager ALTERNATES 6.5.2.3 'All alternate members shallJbe appointed in writing by the SRC [
. Chairman to serve on a_ temporary basis; however, no more than two alternates
-shall participate as' voting members in SRC activities'at am) one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the SRC Chairman to '
provide expert advice .to the SRC.
MEETING FREQUENCY 6.5.2.5 The SRC shall meet at least once per six months.
QUORUM 6.5.2.6 The minimum quorum of the SRC necessary for the performance of the _
l SRC review and audit functions of these Technical Specifications shall consist of the Chairman, or his designated alternate, and at least 3 SRC ,
. members, including alternates.- No more than a minority of the quorum shall have line responsibility for the facility.
' REVIEW-6.5.2.7 The Safety Review Committee shall review: .-
i l a. The safety evaluations for 1) changes to procedures, equipment or I l
systems and 2) tests or experimente completed under the provisions of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
c.' Proposed tests or experiments which involve an unreviewed safety question as defined in section 50.59, 10 CFR.
.d. Proposed changes to Appendix "A" Technical Specifications of this license. :
I G-6 k
~
s V3 e ADMINISTRATIVE CONTROLS - (Cont'd)
================= a=========s-===================================
e. Violstions of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f. Significant deviations from normal and expected performance of facility equipment that affects nuclear safety.
g. All REPORTABLE EVENTS.
h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety.
i. Reports and meeting minutes of the Operations Review Committee.
J. Changes to the Contingency Plan and Plant Security Plan.
k. Changes to ' he Emergency Plan.
1. Changes to the Decommissinning Plan.
AUDITS s
6.5.2.8 Audits of facility activities shall be performed under the cognizance of the SRC. These audits shall encompass:
a. The conformance of facility operation to provisions contained within the Appendix "A" Technical Specifications and applicable license conditions at least once per 24 months.
b. The performance, training and qualifications of the entire facility staff at least once per 24 months.
c. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 12 months,
d. The performance of activities required by the Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.
c. The Emergency Plnn and implementing procedures at least once per 24 months.
f. The Contingency Plan, the Security Plan and implementing proceduses, at least once per 24 months.
g. The Fire Protection Program and implementing procedures at least once per 24 months.
G-7
4 ADMINISTRATIVE CONTROLS - (Cont'd)
=================================================================
h. An independent fire protection and loss prevention program inspection and audit-shall be performed at least once per 24 months >
utilizing either qualified offsite licensee personnel or an outside fire protection firm.
i. ,The Radiological Environmental Monitoring Program and results at.
'least once per 24 months. ;
J. The OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM and implementing procedures at least once per 24 months. :
k. Any other area of facility operation considered appropriate by the :
SRC or the General Manager. ,
AUTHORITY l 6.5.2.9 The SRC shall report to and advise the General Manager on those areas of responsibility listed in Specifications 6.5.2.7 and 6.5.2.8.
REQORDS 6.5.2.10 Records of SRC activities shall be prepared, approved and diatributed as indicated below:
a. Minutes of each SRC meeting shall be prepared, approved and .
forwarded to the General Manager within 20 days following each meeting.
b. Audit repor ts encompassed by Specification 6.5.2.8 above, shall be forwarded to the General Manager and to the management positions responsible for the areas audited within 30 days after completion of the audit.
6.6 PROCEDURES 6.6.1 Written procedures shall he established, implemented and maintained covering the activities referenced below:
a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978. I
b. FUEL HANDLING operations.
-c. Surveillance and test activities of equipment.
d. Security Plan implementation. ;
i
c. Emergency Plan implementation. ;
f. Fire Protection Program implementation.
6-8 E
[
t
ADMINISTRATIVE CONTROLS - (Cont'd)
=================================================================
g. PROCESS CONTROL PROGRAM and 0FFSITE DOSE CALCULATION MANUAL implementation.
6.6.2 Each procedure of Specification 6.6.1, and changes thereto, shall be reviewed by the ORC and approved by the Plant Superintendent prior to implementation and reviewed periodically as set forth in administrative procedures.
6.6.3 Temporary changes to procedures of Specification 6.6.1 may be made provided:
a. The intent of the original procedure is not altered.
b. The change is approved by two members of the facility management staff.
c. The change is documented, reviewed by the ORC and approved by the Plant Superintendent within 30 days of implementation.
6.6.4 SPEC ML PROCEDURE REQUIREMENTS 6.6.4.1 Process control Program (PCEl
a. The PCP shall be maintained on site and will be availnble for NRC review.
b. Licensee initiated changes to the PCP:
1. Shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
o eformation to support the rationale for the change; A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
Documentation of the fact that the change has been reviewed and found acceptable by the ORC.
6.6.4.2 0FFSITE DOSR CAlfUI.ATION MANUAL (ODCM)
The ODCM shall be maintained by the licensee. Changes to the ODCM will be outlined in the Semi-Annual Radioactive Effluent Release Report per Specification 6.8.1.2.
6-9
ADMINISTRATIVE CONTROLS - (Cont'd)
===============2=================================================
This submittal shall contain:
1. Detailed information to support the rationale for the change.
Information submitted should consist of a 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) and
2. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations.
6.7 CONTROL OF MAINTENANCE AND TESTING ACTIVITIES 6.7.1 Maintenance operations and routine tests shall be performed in conformance with these specifications.
6.7.2 Maintenance operations shall be performed as authorized by the Shift Supervisor. Maintenance involving the opening of systems containing radioactive materials shall be conducted under the surveillance of a llealth Physics representative.
6.7.3 Components which have been repaired, replaced, or otherwise subjected to temporary or permanent modification shall be tested in accordance with procedures which are appropriate in view of the nature of the repair, replacement or modification, and in view of the condition of the system, 6.7.4 Key switches shall permit operational, maintenance, and test bypass of the safety instrumentation only with the approval of the Shift Supervisor.
6.8 REPORTING REQUIREMENTS 6,8.1 ROUTINE REPORTS In addition to the applicable reporting requirements of Title 10, code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.
6.8.1.1 Reports required on an annual basis shall be submitted by March 1 of each year and shall include:
a. A tabulation on an annual basis of the number of station, utility and other personnel, including contractors, receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions, 1/e.g., plant operations and surveillance, inservice inspection, routine maintenance, special 1/ This tabulation supplements the requirements of Section 20.407 of 10 CFR Part 20.
6-10
e 2 ADMINISTRATIVE CONTROLS - (Cont'd)
============================================================================= .
maintenance (describe maintenance), waste processing, and fuel handling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling 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.
k A report containing a brief description of any changes, testing and ]
i_ b.
experiments conducted under the criteria of 10 CFR 50.59, including ;
a summary of the safety evaluations of them.
?
r ..
E c. An Annual Radiological Environmental Monitoring Report which shall :
1 E include summarized and tabulated results, including interpretations and analysis of data trends, of environmental sampleu taken during the previous calendar year. In the event that some results are not available for inclusion with the report, the report shall be y submitted noting and explaining the reasons for the missing g results. The missing data shall be submitted as soon as possible
, in a supplementary report.
R The report shall also include the following: a summary description of the Radiological Environmental Monitoring Program; a map of all sampling locations keyed to a table giving distances and directions from the plant, the results of the Interlaboratory Comparison Program, and a discussion of all analyses in which the LLD was not b achievable.
.h G.8.1.2 Semi-Annual Radioacti< e Effluent Release Report Paragraph (a)(2) of Part 50.36a, "Technical Specifications on Effluents from Nuclear Power Reactors," of 10 CFR Part 50 requires that a report
" be made to the Commission within 60 days after January 1 and July 1 of each year. The report shall specify the quantity of each of the principal radionuclides released to unrestricted areas by liquids and gaseous effluents during the previous 6 months. The information I submitted shall be in accordance with Appendix B of Regulatory Guide
, 1.2? (Revision 1) dated June 1974 with data summarized on at least a -
p quarterly basis.
An annual summary of (hourly) meteorological data c. lected over the previous year in accordance with Regulatory Guide 1 41 (Rev. 1) will be
- included with the Semi-Annual Radioactive Effluent Release Report, which is subuitted 60 days after January 1 of each year. This same report shall include an assessment of radiation doses to MEMBERS OF THE PUBLIC from radioactive liquid and gaseous effluents released beyond the EFFLURNT RELEASE BOUNDARY performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM) . This report will contain any changes made to the ODCM during the previous twelve months. _
lI G-11 g-i 4
II r
2
ADMINISTRATIVE CONTROLS - (Cont'd)
=================================================================
6.8.2 SPECIAL REPORTS Special reports shall.be submitted to the Regional Administrator of the Regional Office of the N'c0 within the time period specified for each report.
6.8.3 LICENSEE EVENT REPORTS 6.8.3.1 A Licensee Event Report shall be submitted to the Nuclear Regulatory Commission if any REPORTABLE EVENT of the type listed in Specification
6. 8. 3. 3 occu rs . The report shall meet the requirements of 10 CFR 50.73, with the exception of the events qualifying as REPORTABLE EVENTS.
6.8.3.2 Each REPORTABLE EVENT shall be reviewed by the ORC and the Licensee Event Report shall be submitted to the SRC and General Manger.
6.8.3.3 The following types of events are REPORTABLE EVENTS:
a. Any violation of or condition prohibited by Technical Specifications.
b. Any deviation from Technical Specifications authorized by n 10 CFR 50.54(x).
c. Serious degradation of nuclear fuel.
d. Loss of CONTAINMENT INTEGRITY when it is required to exist.
e. Any event or condition that resulted in the plant being in an unanalyzed condition that significantly corpromised plant safety.
f. Any event or condition that resulted in the plant being in a condition outside its design basis.
g. Any event or condition that resulted in the plant being in a condition not covered by normal or emergency procedures.
h. Any natural phenomenon, other external phenomenon, or event such as fire, toxic gas release, or radioactive release, that posed an actual threat to plant safety or significantly hampered site personnel in the performance of duties necessary for plant safety.
i. Any event or condition that resulted in unplanned and unexpected manual or automatic actuation of the Emergency Diesel Generators or Containment Isolation, 1
G-12 i
L
7 r, -
' ADMINISTRATIVE CONTROLS --(Cont'd)-
+
=================================================================
,) . - Any airborne radioactivity release that exceeded 2 times - the applicable' concentrations of the_ limits specified in Appendix B.
Table . II of 10 CFR Part .20 in unrestricted areas, when averaged over a time period of one hour.
k. .Any liquid effluent release that exceeded 2 times the limiting-combined Maximum Permissible Concentration (MPC) (See Note 1 of Appendix B to 10 CFR Part 20) at the point-of entry into the receiving water (i.e., EFFLUENT RELEASE BOUNDARY) for all radionuclides except tritium and dissolved noble' gases, when-averaged over a time period of one hour.
' 6.8.'4 IPNEDIATE NOTIFICATION ' REQUIREMENTS In lieu of the reporting requirements of 10 CFR 50.72, the events listed in Specification 6.8.3.3 shall be reported to the NRC Operations Center within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Notification of declaration of an Emergency Class listed in the LACBWR Emergency Plan shall be made within I hour.
-6.9 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
ADMINISTRATIVE CONTROLS - (Cont'd)
=======================n=========================================
6.9.1 The following records shall be retained for at least five years:
a. Records and logs of facility operation,
b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c. All REPORTABLE EVENTS submitted to the Coimaission.
d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
e.- Records of changes made to the procedures required by Specification 6.6.1.
f. Records of radioactive shipments.
g. Records of sealed source and fission detector leak tests and results.
6-13
~
hwilNISTRATIVE CONTROLS - (Cont'd)
=================================================================
h. Records of annual physical inventory of all sealed source material of record.
6.9.2 The following records shall be retained for the duration of the LACBWR License:
a. Facility design modification packages,
b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories,
c. Records of radiation exposure for all individuals entering radiation control areas.
d. Records of gaseous and liquid radioactive material released to the environs, and records of analyses required by the Radiological Environmental Monitoring Program.
e. Records of reactor tests and experiments.
f. Records of training and qualification for current members of the facility staff.
g. Records of in-service inspections performed pursuant to these Technical Specifications,
h. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
i. Records of meetings of the ORC and the SRC.
G.10 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
6.11 HIGH RADIATION AREA 6.11.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.203 (c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Special Work Permit (SWP).* 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.
6-14
ADMINISTRATIVE CONTROLS - (Cont'd)
=========================================================================== ,
w
b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them,
c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device and who is responsible for providing positive exposure control over the activities within the area and who will perform periodic radiation surveillance at the frequency which will he established by the Health and Safety Supervisor.
6.11.2 For each area with radiation levels greater than 1000 mRems/hr, the control of Specification 6.11.1 shall be implemented and (1) Each entrance or access point to the area shall be maintained locked except during periods when access to the area is required, with positive control over each individual entry, or (2) Each entrance or access point to the area shall be equipped with a ;
control device which shall energize a conspicuous visible or audible alarm signal in such a manner that the individual entering the high radiation area and the licensee or a supervisor of the activity are made aware of the entry.
Ilealth Physics personnel or personnel escorted by IIcalth Physics personnel shall be exempt from the SWp issuance requirement during the performance of their assigned radiation protection duties, provided they are following plant radiation protection procedures for entry into high radiation areas.
l l
l l
f
{
l G-15

ML20147B779

Text

======================================u==========================

=================================================================

=================================================================

=================================================================

t ===================================================================

=================================================================

=================================================================

==_=============================================================

================================================================

================================================================

=====================xz=========================================

========t=u==2==================================================

====================================== =========================

=================================================================

=================================================================

=================================================================

================================================================-

=================================================================

=============================r.============================.======

=================================================================

=================================================================

===========================-=====================================

=================================================================

=================================================================

=================================================================

=====u===========================================================

=======================================================e=========

======================c===========,==============================

======================================================s==========

=================================================================

=================================================================

=================================================================

=================================================================

===============5=================e===============================

===== ================================c=============:============

====a============================================================

=================================================================

=================================================================

=================================================================

=================================================================

=================================================================

=================================================================

==============e===================================================

=================================================================

==e===================================2==========================

====n============================================================

====,============================================================

===ers== == a22:e======== r===nusmermaw=r=== scrars=====neestar==s=====

================ =======z=============================2==========

===================================================================== -

=================================================================

=================================================================

=================================================================

================================================================

=================================================================

=================================================================

= ===============================================================

=================================================================

=================================================================

=======================================r:===========:=============

=================================================================

========================================================a========

===================e==========================================n==

=================================================================

=================================================================

=================================================================

=================================================================

========================================================se=======

=======:=======================================================

================= a=========s-===================================

=================================================================

=================================================================

===============2=================================================

=================================================================

=================================================================

=======================n=========================================

=================================================================

Navigation menu