ML20081D105

Proposed Tech Spec Changes,Incorporating Radiological Effluent Requirements of 10CFR50,App I

ML20081D105
Person / Time
Site:	Peach Bottom
Issue date:	03/07/1984
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20081D092	List: ML20081D090 ML20081D096 ML20081D105
References
NUDOCS 8403150143
Download: ML20081D105 (82)

Text

__ ___. _ _.

j PBAPS TABLE OF CONTENTS (Cont'd)

Pace SURVEILLAFCE LIMITING CONDITIONS FOR OPERATION REOUIRFFFNT l

3.6 PRIMARY SYSTEM BOUNDARY

'4.6 143 A.

Thermal and Pressurization Limitations A

14 3 B.

Coolant Chemistry B

145 C.

Coolant Leakage C

146 D.

Safety and Relief Valves D

147 E.

Jet Pumps E

148 F.

Jet Pump Flow Mismatch F

148 G.

Structural Integrity G

149 i

3.7 CONTAINMENT SYSTEMS 4.7 165 A.

Primary containment A

IG5

~

B.

Standby Gas Treatment System B

175 C.

Secondary Containment C

'176 D.

Primary Containment Isolation Valves D

177 l

3.8 RADIOACTIVE MATERIALS 4.8 203 A.

General A

203 B.

Liquid Ef fluents B

204 C.

Gaseous Ef fluents C

20R D.

40 CFR 190 D

236 E.

Radiological Environmental Monitoring E

236a-1 i

3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 21 7 A.

Auxiliary Electrical Fouipment A

21 7 B.

Operation with Inoperable Eouipment B

219 C.

Emergency Service Water System C

221 3.10 CORE 4.10 225 A.

Refueling Interlocks A

225 B.

Core Monitoring B

227 i

C.

Spent Fuel Pool Water Level C

228 j

D.

Heavy Loads Over Spent Fuel D

228 E.

Spent Fuel Decay Time E

228 f

3.11 ADDITIONAL SAFETY RELATED PLANT CAPABILITIES 4.11 233 j

A.

Main Control Room Ventilation A

233 B.

Alternate Heat Sink Facility B

234 C.

Emergency Shutdown Control Panel C

234 11

.i B403150143 840307 PDR ADOCK 05000277 p

PDR

1 PBAPS TABLE OF CONTENTS (Cont'd)

Pace LIMITING CONDITIONS FOR OPERATION SUFVEILLANCE REOUIREMENT 3.12 RIVER LEVEL 4.12 237 A.

High River Water Level B.

Low River Water Level A

237 C.

Level Instrumentation B

237 C

238 3.13 MISCELLANEOUS RADIOACTIVE MATERIALS SOURCE l

4.13 240a 3.14 FIRE PROTECTION 4.14 240c A.

Water Fire Protection System B.

A CO2 Fire Protection System 240c C.

Fire Detection B

240g D.

Fire Barrier Penetrations C

2401 D

240j 3.15 SEISMIC MONITORING INSTRUMENTATION 4.15 240n 5.0 MAJOR DESIGN FEATURES 241 6.0 ADMINISTRATIVE 00NTROLS 243 6.1 Responsi bility 6.2 Organization 243 6.3 Facility Staf f Qualifications 24 3 6.4 Trainino 246 6.5 l

Review and Audit 246 6.6 Reportable Occurrence Action 246 6.7 l

Safety Limit Violation 253 6.8 Procedures 253 i

6.9 Reporting Peauirements 253 1

6.10 Record Retention 254 6.11 Radiation Protection Program 260 6.12 Fire Protection Inspections 26]

6.13 Hioh Radiation Area 261 6.14 Integrity of Systems Outside Containment 262 6.15 Iodine Monitoring 263 6.16 Environmental Qualifiestion 264 6.17 Of fsite Dose Calculation Manual 264 6.18 Major Changes to Radioactive Waste 265 Treatment Systems 266 111

=

l i

PBAPS 1

l Unit 2 LIST OF FIGURES i

Figure Title Page l

t 1.1-1 APRM Flow Bias Scram Relationship 16 s

To Normal Operating Conditions 4.1.1 Instrument Test Interval Determina-55 tion Curves 4.2.2 Probability of System Unavailability 98 Vs. Test Interval 3.4.1 Required Volume and Concentration of 122 Standby Liquid Control System Solution

-3.4.2 Required Volume and Concentration of 123 Standby Liould Control System Solution

• Figures 3.5.1.A and 3.5.1.B (7X7 Fuel) deleted.

(PB2 Cycle 5-all 8X8 core) 3.5.1.C MAPLHGR Vs. Planar Average Exposure, 142b Unit 2, 8x8 f uel, Type H - 80 mil &

100 mil i

3.5.1.D MAPLHGR Vs. Planar Average Exposure 142c Unit 2, 8x8 Fuel, Type L 3.5.1.E Kf Factor Vs. Core Flow 142d 3.5.1.F MAPLHGR Vs. Planar Average Exposure, 142e Unit 2, 8x8 LTA Fuel,100 mil channels 3.5.1.G MAPLHGR Vs. Planar Average Exposure, 142f 4

Unit 2, 8X8R Fuel, Type BDRB284, 100 mil channels 3.5.1.H MAPLHGP Vs. Planar Average Exposure, 1429 Unit 2, P 8X8R Fuel, Type P8DRB285, 100 mil channels 3.5.1.1 MAPLHGR Vs. Planar Average Exposure, 142h Unit 2, P BXBR Fuel, Type P8DRB284 H, 80 mil & 100 mil channel & 120 mil channel s

-iv-

.l

9 i

PBAPS Unit 2 LIST OF FIGUPES Floure Titl e Pace 3.6.1 Minimum Temperature for Pressure 164 Tests such as required by Section XI 3.6.2 Minimum " emperature for Mechanical 164a Heatup r e Cooldown following Nuclear Shutdown 3.6.3 Minimum Temperature for Core Operation 164b (Criti cality) 3.6.4 Transition Temperature Shif t vs.

164c Fluence l

3.8.1 Site Boundary and Ef fluent Release 216e Points s

6.2-1 Management Organization Chart 244 6.2-2 Organization for Conduct of Plant 245 Operations iva l

1

'rBAPS Unit 3

_ LIST OF FIGURES Floure Title Pace 1.1-1 APRM Flow Bias Scram Relationship To Normal Operating Conditions 16 4.1.1

)

Instrument Test Interval Determina-tion Curves 55 4.2.2 Probability of System Unavailability Vs. Test Interval 98 3.4.1 Required Volume and Concentration of Standby Liquid Control System Solution 122 3.4.2 Reauired Temperature vs. Concentration for Standby Liquid Control System 123 3.5.K.1 MCPR Operating Limit Vs. Tau 8X8 and 8XBR Fuel 142 3.5.K.2 MCPR Operating Limit Vs. Tau, P 8X8R Fuel 142a 3.5.1.C MAPLHGR Vs. Planar Average Exposure, Unit 2*, 8x8 Fuel, Type H 142b 3.5.1.D MAPLHGR Vs. Planar Average Exposure Unit 3, Ox8 Fuel, Type L 142c 3.5.1.E Kf Factor Vs. Core Flow 142d 3.5.1.F MAPLHGR Vs. Planar Average Fxposure, Unit 3, 8x8 PTA Fuel 14?e 3.5.1.G MAPLHGP Vs. Planar Average Exposure Unit 3, 8x8R Fuel 142f 3.5.1.H MAPLHGR Vs. Planar Average Exposure, Unit 3, P8x8R Fuel (P8DRB284H )

142g 3.5.1.I MAPLHGR Vs. Planar Average Exposure, Unit 3, P8x8R Fuel 142h (P8DRB299) 3.5.1.J MAPLHGR Vs. Planar Average Exposure, Unit 3 P8x8R Fuel 142i (Generic) iv l

PBAPS Unit 3 LIST OF FIGURES

_ Figure Title Page 3.6.1 Minimum Temperature for Pressure 164 Tests Such as Required by Section XI 3.6.2 Minimum Temperature for Mechanical 164a Heatup or Cooldown Following Nuclear Shutdown 3.6.3 Minimum Temperature for Core Operation (Criticality) 164b 3.6.4 Transition Temperature Shift vs. Fluence 164c 3.8.1 Site Boundary & Effluent Release Points l

216e 6.2-1 Management Organization Chart 244 6.2-2 Organization for Conduct of Plant 245 Operations I

iva i

23AP3 Unit 2 LIST OF TABLES Tabl e Title Pace 4.2.B Minimum Test and Calibration Frequency 81 for CSCS 4.2.C Minimum Test and Calibration Frecuency 83 for Control Rod Blocks Actuation 4.2.D Minimum Test and Calibration Fr sauency B4 for Radiation Monitoring Systems 4.2.E Minimum Test and Calibration Frecuency 85 for Drywell Leak Detection 4.2.F Minimum Test and Calibration Frecuency 86 for Surveillance Instrumentation 4.2.G Minimum Test and Calibration Frecuency 88 for Recirculation Pump Trip 3.5.K.2 Operating Limit MCPR Values for 1336 Various Core Exposures 3.5.K.3 Operating Limit MCPR Values for 133e Various Core Exposures 4.6.1 In-Service Inspection Program for Peach 150 Bottom Uni ts 2 and 3 5.7.2 Primary Containment Isolation Valves 179 3.7.2 Testable Penetrations With Double 184 0-Ring Seals 3.7.3 Testable Penetrations with Testable IR4 Bellows 3.7.4 Primary Containment Testable Isolation IP5 Val ves 4.8.1 Radioactive Liquid Waste Sampling and 216b-1 Anal ysis 4.8.2 Radioactive Gaseous Waste Sampling and 216c-1 Analysis 4.8.3.a Radiological Environmental Monitoring 216d-1 Program 4.8.3.b Reporting Levels for Radioactivity 216d-5 by Concentrations in Environmental Sampl e vi

PBArm Unit 2 LIST OF TABLES Table Title Pace 4.8.3.c Maximum Values for Minimum Detectable 2166-6 Levels of Activity 3.11.D.1 Safety Related Simck Suppressors 234d 3.14.C.1 Fire Detectors 240k l

via i

l

PBAPS Unit 3 LIST OF TABLES Tabl e Titl e Pace 4.2.B Minimum Test and Calibration Frecuency 81 for CSCS 4.2.C Minimum Test and Calibration Frecuency 83 for Control Rod Blocks Actuation 4.2.D Minimum Test and Calibration Frequency 84 for Radiation Monitoring Systems 4.2.E Minimum Test and Calibration Frecuency 85 for Drywell Leak Detection 4.2.F Minimum Test and Calibration Frequency 86 for Surveillance Instrumentation 4.2.G Minimum Test and Calibration Frecuency 88 for Recirculation Pump Trip 3.5.K.2 Operating Limit MCPR Values for 133d Various Core Exposures 3.5.K.3 Operating Limit MCPR Values for 133e i

Various Core Exposures 4.6.1 In-Service Inspection Program for Peach 150 Dottom Units 2 and 3 3.7.1 Primary Containment Isolation Valves 179 3.7.2 Testable Penetrations With Double 184 0-Ring Seals 3.7.3 Testable Penetrations With Testable 184 Bellows 3.7.4 Primary Containment Testable Isolation 185 l'

Valves 4.8.1 Radioactive Liould Waste Sampling and 216b-1 Analysis 4.8.2 Radioactive Gaseous Waste Sampling and 216c-1 Analysis 4.8.3.a Radiological Environmental Monitoring 216d-l Program 4.8.3.b Reporting Levels for Radioactivity 216d-5 Concentrations in Environmental Samples vi

,e--w-

-ew.

--ee,--m c--,-

-e--a

--p,,,--c,

-,.m

~--,-m---

s.-o-,

t PBAPS Unit 3 LIST OF TABLES Table Title Page 4.8.3.c Maximum Values for Minimum Detectable 216d-6 Levels of Activity 3.11.D.1 Safety Related Shock Suppressors 234d 3.14.C.1 Fire Detectors 240k f

i l

l 1

e I

l t

via l

l I

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

I PBAPS 1.0 DEFINITIONS The succeeding f recuently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.

Alteration of the Reactor Core - The act of moving any component in the region above the core support plate, below the upper grid and within the shroud.

Normal control rod movement with the control drive hydraulic system is not defined as a core alteration.

Normal movement of in-core instrumentation and the traversing in-core probe is not defined as a. core alteration.

Channel - A -channel is.an arrangement of a sensor and associated components used to evaluate plant variables and produce discrete b

outputs used in logic.

A channel terminates and loses its identity where individual channel outputs are combined in logic.

Cold Condition - Reactor coolant temperature ecual to or less than 212 F.

Cold Shutdown - The reactor is in the shutdown mode, the reactor coolant temperature equal to or less than 212 F, and the reactor vessel is vented to atmosphere.

Critical Power Ratio (CPR) - The critical power ratio is the ratio of that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as calculated by application of the GEXL correlation.

(Reference NEDO-10958).

4 Dose Equivalent I-131 - That concentration of I-131 ( Ci/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133 I-134, and I-135 actually present. l

PBAPS 1.0 DEFINITIONS (Con t ' d )

Enoineered Safeonard - An engineered safeguard is a safety system the actions of which are essential to a safety action reouired in response to accidents.

Traction of Limitino Power Density (FLPD) - The ratio of the Tinear heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type.

Functional Tests - A functional test is the manual operation or initiation of a system, subsystem, or component to verify that it functions within design tolerances (e.g., the manual start of a core spray pump to verify that it runs and that it pumps the reouired volume of water).

Gaseous Radwaste Treatment System - Any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

Hot Shutdown - The reactor is in the shutdown mode and the reactor coolant temperature greater than 212 F.

Hot Standby condition - Hot Standby Condition means operation with coolant temperature greater then 212 F, system pressure less than 1055 psig, and the mode switch in the Startup/ Hot Standby posi ti on.

The main steam isolation valves may be opened to provide steam to the reactor feed pumps.

Immediate - Immediate means that the reouired action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the recuired action.

i

2

PBAPS 1.0 DEFINITIONS (Cont'd)

Instrument or Channel Calibration - An instrument or channel calibration means the adjustment of an instrument or channel signal output so that it corresponds, within acceptable range, and accuracy, to u known value(s) of the parameter which the instrument or channel monitors.

The known value of the parameter shall be injected into the channel or instrument as close to the primary sensor as practicable.

Instrument or Channel Check - An instrument or channel check is a cualitative determination of acceptable operability by observation of instrument or channel behavior during operation.

This determination shall include, where possible, comparison of the instrument or channel with other independent instruments measuring the same variable.

Instrument or Channel Functional Test - An instrument or channel functional test means the injection of a simulated signal into the channel or instrument as close to the primary sensor as practicabic to verify the proper instrument channel response, alarm and/or initiating action.

Limitino Conditions for Operations (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility.

When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.

Limitina Safety System Setting (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded.

The region between the safety limit and these settings represents margin with normal operation lying below these settings.

The margin has been established so that with proper operation of the instrumentation, the safety limits will never be exceeded.

Logic - A logic is an arrangement of relays, contacts and other components that produces a decision output. (

1 I

l

PBAPS 1.0 DEFINITIONS (Cont'd)

)

(a) Initiatina - A logic that receives signals from channels and produces decision outputs to the actuation logic.

(b) Actuation - A logic that receives signals (either f rom initiation logic or channels) and produces decision outputs to accomplish a protective action.

Logic System Functional Test - A logic system functional test means a test of all relays and contacts of a logic circuit to insure all components are operable per desi Where practicable, action will go to completien; gn intent.

i.e.,

pumps will be started and valves operated.

" Maximum Fraction of Limitino Power Density (MFLPD) - The Maximum

' Fraction of Limiting Power Density (MFLPD) is the highest value existing in the core of the Fraction of Limiting Power Density (FLPD).

MEMBERS OF THE PUBLIC - Members of the public shall include all persons who are not occupationally associated with the plant.

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 to make deliveries.

This category does incJ ude persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

Minimum Critical Power Ratio (MCPR) - The minimum in-core critical power ratio corresponding to the most limiting fuel assembly in the core.

Mode of Operation - A reactor mode switch selects the proper interlocks for the operational status of the unit.

The following are the modes and interlocks provided: Refuel Mode, Run Mode, Shutdown Mode, Startup/ Hot Standby Mode.

.,,....,...- -.- n

sL PBAPS 1.0 DEFINITIONS (Cont'd)

Offsite Dose Calculation Manual - Contains the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and describes the environmental radiological monitoring program.

OPERABLE - OPERABILITY - A system, subsystem, train, component, or device is OPERABLE or has OPERABILITY when it is capable of performing its specified function and all instrumentation, controls, normal and emergency electrical power sources, cooling or seal water supplies, lubrication systems, and other auxiliary equipment that are recuired for the system, subsystem, train, component, or device to perform its function are also capable of performing their related support function.

ratino - Operating means that a system or component is

. forming its intended functions in its required manner.

)perating Cycle - Interval between the end of one refueling sutage f or a particular unit and the end of the next subsecuent

'efueling outage for the same unit.

'rimary Containment Inteority - Primary containment integrity seans that the drywell and pressure suppression chamber are ntact an 3 all of the following conditions are satisfied:

1.

All non-automatic containment isolation valves on lines connected to the reactor coolant system or containment which are not recuired to be open during accident conditions are cl os ed.

These valves may be opened to perform necessary operational activities.

s U

2.

At least one door in each airlock is closed and sealed.

3.

All automatic containment isolation valves are operable or deactivated in the isolated position.

4.

All blind flanges and manways are closed. 4

~

_ _ ~

l PBAPS 1.0 DEFINITIONS (Cont'd)

Protective Action - An action initiated by the protection system when a limit is reached.

A protective action can be at a channel or system level.

l Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

Purae - Puraina - Purge or Purging is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or, other operating condition, in such a manner that replacement air or gas is

- reauired to purify the confinement.

Rated Power - Rated power refers to operation at a reactor power of 3,293 MWt; this is also termed 100 percent power and is the maximum power level authorized by the operating license ' Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters I

when the reacter is at rated power. '

Reactor Power Operation - Reactor power operation is any operation with the mode switch in the "Startup" or "Run" position with the reactor critical and above 14 rated power.

Reactor vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

Refuel Mode - With the mode switch in the refuel position, the reactor is shutdown and interlocks are established so that only one control rod may *oe withdrawn.

Ref telino Outaae - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling.

For the purpose of designatina frecuency of testing and surveillance, a refueling outage shall mean a regularly scheduled outage however, where such outages occur within 8 months of the completion of the previous refueline

PBAF5 1.0 DEFINITIONS (Cont'd) outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

Run Mode - In this mode the reactor system pressure is at or above 850 psig and the reactor protection system is energized with APRM protection (excluding the 15% high flux trip) and RBM interlocks in service.

Safety Limit - The safety limits are limits below which the reasonable maintenance of the cladding and primary systems are assured.

Exceeding such a limit requires unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation.

Operation beyond such a limit may not in itself result in serious consequences, but it indicates an operational deficiency euhject to regulatory review.

Secondary containment Integrity - Secondary containment integrity means that the reactor building is intact and the following conditions are met:

1.

At least one door in each access opening is closed.

2.

The standby gas treatment is operable.

3.

All Reactor Building ventilation system automatic isolation valves are operable or deactivated in the isolation position.

Shutdown - The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations are being performed.

Shutdown Mode - Placing the mode switch to the shutdown position l

initiates a reactor scram and power to the control rod drives is removed.

After a short time period signal is removed allowing a scram re(about 10 sec), the scram set and restoring the normal valve lineup in the control rod drive hydraulic system; also, the main steam line isolation scram and main condenser low vacuum scram are bypassed if reactor vessel pressure is below 1055 psig.

I PBAPS 1.0 DEFINITIONS (Cont'd)

Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circui t in question.

SITE BOUNDARY - That line beyond which the land is not owned, lensed or otherwise controlled by licensee.

o Source Check - A source check shall be the qualitative assestment of channel response when the channel. sensor is exposed to a radioactive source.

Startup/ Hot Standby Mode - In this mode the reactor protection scram trips, initiated by condenser low vacuum and main steam line isolation valve closure, are bypassed when reactor pressure is less than 1055 psig, the reactor protection system is energized with IRM neutron monitoring system trip, the APRM 15%

high flux trip, and control rod withdrawal inte,rlocks in service.

This is of ten ref erred to as just Startup Mode.

This is intended to imply the Startup/ Hot Standby position of the mode switch.

Surveillance Frecuency - Periodic surveillance tests,

checks, calibrations, and examinations shall be performed within the specified surveillance intervals.

The operating cycle interval as pertaining to instrument and electrical surveillance shall not exceed 18 months.

These specified time intervals may be exceeded by 25%.

In cases where the elapsed interval has exceeded 100% of the specified interval, the next surveillance interval shall commence at the end of the original specified interval.

Surveillance tests are not required on systems or parts of the systems that are not required to be operable or are tripped.

If tests are missed on parts not required to be operable or are tripped, then they shall be performed prior to returning the system to an operable status.

Transition Boiling - Transition boiling means the boiling regime between nucleate and film boiling.

Transition boiling is the regime in which both nucleate and film boiling occur i

intermittently with neither type being completely stable,.

Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary eauipment required to initiate 1 l

...: a 1.0 DEFINITIONS (Cont'd) l action to accomplish a protective trip function.

A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action.

Initiation or protective action may require the tripping of a single trip system or the coincident tripping of two trip syste ms.

Unrestricted Area - Any area for which access control is not required for purposes of protection of individuals from exposure to radiation.

Ventilation Exhaust Treatment System - A ventilation exheust treatment system is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents.

Ventino - The controlled process of discharging air or gas from a con finement to maintain operating conditions, such that replacement air or gas is not provided or required during venting.

(

1

-Ba-

PBAPS LIMITING CDNDITIONS FOR OPERATION SUFVEILLANCE RFOUIRhMENTS

[

C.

Control Rod Block Actuation C.

Control Rod Block 5ctuation 1.

The limiting conditions of Instrumentatien shall be operation for the instru-f unctionally. te st ed, cali- -

mentation that initiates brated and checked as indi-control rod blocks are given cated in Table 4.'2.C.

In Tabl e 3. 2.C.

System logic shall be func-2.

The minimum number of oper-tionally tested as indica-able instrument channels ted in Table 4. 2.C.

specified in Table 3.2.C for the Rod Block Monitor may be reduced by one in one of the trip systems for s

maintenance and/or testing, provided that this condi-tion does not.last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty

/

day period.

as. - -

e-Pm '3

,'t.*

A b

4 esa,_,

j i

s s

Y sdas y

y..

i

\\

'N's sZ C

W]

s. s

.s s

sf

..^

y' > ~

~

, '\\

~;r 4

s u

~~

.m s

n

.x s

- s,

\\

_l,%.

• w.

PBAPS

, LIMITING CONDITIONS FOR OPERATION SUFVETLLANCE REOUIRTMFFTS 3.2.D. Radiati'an Monitoring 4.2.D.

Radiation Monitorino Systems-Isolation and

_ Initiation Functions Systems-Isolati on and i..

Ini tiati on Functions

\\

1.

Reactor Building Isolation 1.

Reactor Buildino Isolation l

and ' Standby Gas Treatment System

~

~

and Standby Gas Treatment System The limiting conditions Instrumentation shall be

/

for operation c:6 given in functionally tested, cali-Tabl e 3. 2.D.

brated and checked as indi-cated in Table 4.2.D.

System logic shall be func-

~

tionally tested as indica-ted in Table 4. 2.D.

f,,

prwell Leak Detection E.

Drywell Leak Detection The limiting Manditions of Instrumentation shall be operation for'the instru-calibrated and checked as mentation.that monitors drywell leak detection are indicated in Table 4.2.E.

given in Tabl e 3.2.E.

y

=

s

'~

-[r

"~

~^-r gm s

s 4R M'.

j_

4%

s e

N 9

l f

e 9.

L s

s s

l l

TABLE 3.2.D RADIATION MONITORING SYSTEMS THAT INITIATE AND/OR ISOLATE SYSTEMS t

Minimum No. of Operable In strument,'

No of Instrument

' Channel s Trip Function Trip Level Setting by Design (2)

Channels Provided Action i

)

p 7

Kefuel Area Exhaust Monitor Upecale, <16 pr/hr 4 Inst. Channel s A or B 2

Reactor Building Area

. Upscale, (16 mr/hr 4 Inst. Channels B

Exhaust Monitors l

g

., f'tiOTES FOP TABLE 3. 2.D

1. Whenever the systems arc required to be operable, there dhall be two operable,or tripped

. ~instrtunent channel: ' per trip system. If this cannot be met, the indicated. action shall be taken.

' 2. Action Cease operatkm of the refueling eouipment.

~'

A.

B. Isolate secondary containment and start the standby gas treatment system.-

l}

a a

1 En "T

/

TABLE 4.2.D MINIMJM TEST & CALIBRATION FREOUENCY FOR RADIATION MONITORING SYS

_ Instrument Channels Instrument Functional Test Instrument i

Calibration "Theck (2)

1) Refuel Area Exhaust (1)

Once/3 months Once/ day Moni tors - Upscale

2) Reactor Building Area (1)

Once/3 months Once/ day Logic System Functional Tcat (4) (6)

Frecuency

1) Reactor Building Isolation l

Once/6 months 7

2) Standby Gas Treatment Once/6 months System Actuation t

4, a

j i

S n

PBAPS 3.2 BASES (Cont'd) i The APRM rod block function is flow biased and prevents a significa t reduction in MCPR, APRM provides gross core protectionsespecially during operation at reduced flo n

The increase from withdrawal of control rods in the normal withdrawali.e.,

sequence.

the fuel cladding integrity safety limit.The trips are set so that MCPR The RBM rod block function provides local protectio core, for a single rod withdrawal error f rom a limiting control rod e pattern.

The IRM rod block function provides local as well as gross core protection.

.than _a factor of 10 above the indicated level.The scaling arrangem

> The.downscale indication on an APRM or IRM is an indication instrument has failed or the instrument is not sensitive enou either case the instrument will not respond to changes in the control g.

In

{

rod motion and thus, control trips are set at 2.5 indicated on scale. rod motion is prevented. The downscale have only one logic channel and are not required fo ento comparator must be bypassed when operating with one recirculation The flow water pump.

The refueling interlocks also operate one logic channel required for safety only when the mode switch is in the refueling

, and are position.

For effective emergency core cooling for small pipe breaks enough to allow either core spray or LPCI to operate in the HPCI automatic pressure relief function is provided as a backup to the HPCI The in the event the HPCI does not provide this function when neces and minimize spurious operation.

sary The trip settings given in the specification are adequate to assure the above criteria are met specification preserves the effectiveness of the system during periods The of maintenance, testing, or calibration, and also minimizes the risk of inadvertent' operation; service.

i.e., only one instrument channel out of 1

PBAPS 1

3.2 BASES (Cont'd)

Four sets of two radiation monitors are provided which initiate the Reactor Building Isolation function and operation of the standby gas treatment system.

Four instrument channels monitor the radiation from the refueling area ventilation exhaust ducts and four instrument channels monitor the building ventilation below the refueling floor.

Each set of the instrument channels is arranged in a 1 out of 2 twice trip logic.

Trip settings of <16 mr/hr for the monitors in the refueling area ventilation exhaust ducts are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the standby gas treatment system.

Flow integrators are used to record the integrated flow of,licpid from the drywell sumps.

The alarm unit in each integrator is set to t

I annunciate before the values specified in Specification 3.6.C are exceeded.

An air sampling system is also provided to detect leakage inside the primary containment.

For.each parameter monitored, as listed in Table 3.2 F, there are two (2) channels of instrumentaticn.

By comparing readings between the two (2) channels, a near continuous surveillance of instrument performance is available.

Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument readings.

The recirculation pump trip has been added at the suggestion of ACRS as a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient.

The response of the plant to this postulated event fall within the envelope of study i

events given in, General Electric Company Topical Report, NEDO-10439, dated March,1971.

In the event of a loss of the reactor building ventilation system, t

radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200 degrees F.

Restoration of the main i

steam line tunnel ventilation flow momentarily exposes the temperature sencors to high gas temperatures.

The momentary temperature increase can cause an unnecessary main ateam line isolation and reactor scram.

Permission is provided to increase the temperature trip setpoint to 250 degrees F for 30 minutes during restoration of the ventilation system to avoid unnecessary plant transient.

l --.. -

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS I

3.8 Radioactive Materials 4.8 _ Radioactive Materials Appli cabili ty Applicability Applies to the radioactive ef fluents f rom the plant.

Applies to the periodic monit-oring and recording of radio-active effluents.

Objective Objective To assure that radioactive material is not released to To ascertain that radioactive the environment in an uncon-releases are as low as trolled manner and to assure reasonably achievable and that any material released within allowable val ues.

is kept as low as reasonably achievable and, in any event.

is within the limits of 10 CFR 20.

> Speci fication Speci fi cati on A.

General A.

General It is expected that releases of radioactive material in Operating procedures shall effluents will be kept at small be developed and used, and eouipment which has been f ractions of the limi ts speci-installed to maintain control fled in Section 20.106 of 10 CFR Part 20 and as further over radioactive materials specified in these Technical in gaseous and liauld effluents Specifications. At the same produced during normal rea ctor operations, inc3 uding ex time the licensee is permitted the flexibility of operation, operational occurrences,pected shall compatible with considerations be maintained and used, to keep of health and safety, to assure levels of radioactive material that the public is provided a in ef fluents released to areas dependable source of power at and beyond the SITE BOUNDAPY even under unusual operating as low as reasonably achievable.

conditions which may tem-porarily result in releases higher that such small fraction s,

but still within the limits specified in Specifications 3.8.B.1 and 3.8.C.1, and in Section 20.106 of 10 CFR Part

20. It is expected that in using this operational flex-ibility under unusual operating conditions the licensee will 203 i

l

~^

PBAPS

_ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RFOUIREMFNTS exert his best efforts to keep levels of radioactive material in effluents as Iow as reasonably achievable.

i 3.8.B Liouid Radwaste Effluents 4.8.B Liquid Radwaste Effluents 1.

The concentration of radie-active material released to la. Facility records shall be

~

areas at and beyond the SITE maintained of the radio-BOUNDARY (See Figure 3.8.1) active concentrations shall be limited to the con-and volume before dilution centration specified in of each batch of liould 10 CFR 20 Appendix B, Table II, ef fluent released, and of Column 2 for radionue11 des the average dilution flow other than noble gases and and length of time over 2x10~' uCi/ml total activit which each discharge concentration for all dis y

occurred.

C solved or entrained noble lb. Prior to release of each

, gases. With -the concentration batch of liould effluent, of radioactive material re-a sample shall be taken leased to areas at and beyond from that batch and analyzed the SITE BOUNDARY exceeding for the concentration of these limits, without delay each significant gamma energy decrease the release rate peak. The release rate of radioactive materials shall be based on the and/or increase the dilution circulating water flow flow rate to restore the rate at the tire of

/

concentration to within the discharge.

limi ts.

Ic. Radioactive licuid waste sampling and activity analysic shall be performed in accordance with Table 4.8.1.

2.

The dose or dose commit-2.

Cumulative dose contri-ment to a MEMBER OF THE PUBLIC from radioactive butions shall be determined materials in liould effluent in accordance with the releases from the two methodology and parameters

(

reactors at the site to in the Offsite Dose areas at and beyond the SITF Calculational Manual (ODCM )

BOUNDARY (see Figure 3.8.1) at least once per month.

shall be limited to:

a. During any calendar quarter to < 3.0 mrem to the totaT body and to

< 10.0 arem to any organ,

and,

b. During any calendar year 204

1 PBAPS LIMITING CDNDITIONS FOR OPERATION SURVEILLANCE FFCCIRFPFNTS to < 6.0 mrem to the total body and to

< 20.0 mrem to any organ.

When the calculated dose f rom the release of radioactive materials in liquid effluents exceeds any of the above limits, prepare and submit to the Commission within 21 torking days, pursuant to Speci fication 6.9.3, a Special Report which identifies the causes for exceeding the limits and corrective actions that have been taken to reduce the releases of radioactive materials in liquid effluents and proposed corrective actions to be taken to assure that subsequent releases are within the limits.

This Special Peport sha3] al so include (1) results of radio-logical analyses of the drinking water source and (2) the radiological impact on the potentially affected drinking water supplies with regard to 40 CFR 141, Safe Drinking Water Act. Reactor shut-down is not reauired.

f 3.

During release of radioactive 3a. The licuid radwaste ef-wastes, the following fluents radiation monitor conditions shall be met:

a. The minimum dilution shall be calibrated water reouired to every 12 months with a satisfy 3.8.B.1 shall known radioactive source be met.

positioned in a reproducible

b. The gross activity geometry with respect to the monitor and flow monitor sensor and every cuarter on the waste effluent by means of a source check. Additionally, an line shall be operable instrument functional test except as specified in shall be performed every 205

~

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMFNTS j

3.8.B.3.d and 3.8.B.3.e.

below.

month and an instrument

c. The effluent control check shall be performed monitor shall be set in every day during release.

accordance with the Functional test shall methodology and parameters demonstrate operability in the ODCM to alarm and of the radwaste discharae automatically close the automatic isolation valve, waste discharge valve and control room an-prior to exceeding the nunciation if any of the limits specified in following conditions exist:

3.8.B.1 above,

1. Instrument indicates

d. From and after the date measured levels above that the gross activity the alarm / trip set-monitor on the waste point.

effluent line is made

2. Instrument indicates or found to be inoperable a downscale failure.

for any reason effluent rel eases may co,ntinue 3b. The liquid effluent flow only if best efforts monitor shall be cali-are taken to make such brated every 12 months.

monitor operable, Additionally, an instru-provided that prior ment check shall be to initiating a performed every day rel ease :

during release.

1.

At least two in-dependent samples of the tank's contents are anal yz ed, and r

2.

At least two technically l

qualified members of the Facility Staff independently verify the release rate cal-culation and discharge line valving.

e. From and after the date that the flow monitor on the waste effluent line is,made or found to be inoperable for any reason, ef fl uent releases via this pathway may continue only if best efforts are taken to make such monitor operable, i

provided that the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.

Pump performance curves 206 1

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREb'EfCS may be used to estimate fl ow.

f. If the requirements of 3.8.B.3.a, 3.8.B.3.b, 3.8.B.3.c, 3.8.B.3.d, or 3.8.B. ?. e cannot be met,

suspend release of radio-active effluents via this pathway.

g. With less than the minimum number of radioactive liould radwaste monitors OPERABLE exert best efforts to return the instruments to OPERABLE status within 30 days and if unsuccessful explain in the next Semi-annual Radioactive Ef fluent

, Release Report why tha in-operability was not cor-rected in a timely manner.

4.

All licuids shall be 4a. Doses due to licuid processed throuch either ef fluent releases to the waste collector filter areas at and beyond the

.and demineralizer, the SITE BOUNDARY shall be fJ oor drain filter, or the fuel pool filter deminer-projected once per month in accordance with the alizer as appropriate prior methodology and parameters i

to their discharge when in the ODCM.

i the projected dose due i

to the licuid ef fluent 4b. The waste collector releases to unrestricted filter and demineralizer areas, when averaged over and the floor drain filter any month, exceeds 0.12 shall be demonstrated mrem to the total body or O.4 mrem to any organ operable once per quarter, I

from the two reactors at unless utilized to process liquid waste during the the site. With liquid previous 13 weeks, by waste being discharged without treatment as analyzing the liauid i

processed through the reauired above, prepare and submit to the Com-appropriate eouipment to determine that it meets mission within 21 working the reauirements of days pursuant to Specifi-Specification 3.8.B.1.

cation 6.9.3, a Special The fuel pool filter Peport which includes the followino informations demineralizer is. exempt from this reouirement

a. Explanation of Why since it is 'an alternate licuid radwaste was treatment system which is 207

PBAFS LIMITING CONDITIONS FOR OPERATION SURVEILLANCT RFOUIREMFNTS being discharged without treatment, not routinely used to identi fication of process lionids for discharge.

any inoperable equipment or sub-systems and the reason for the inoperability,

b. Action taken to restore the in-operable equipment to operable status,

c. Action taken to prevent a recurrence.

Reactor shutdown is not required.

3.8.C Gaseous Ef fluents 4.8.C Gaseous Effluents 1.

The dose rate in areas at and beyond the SITE la. The dose rate due to noble BOUNDARY (see Figure gases in gaseous effluents 3.8.1) due to radioactive shall be determined to be materials in gaseous within the limits in effluents released from the accordance with the methods and procedures of the

.two reactors at the site shall ODCM.

be limited to the following:

a. The dose rate for

(

noble gases shall be Ib. The dose rate due to limited to < 500 iodine-131, iodine-133, mrem /yr to the total tritium, and all radio-body and < 3000 mrem /yr nuclides in particulate to the skin.

form with half lives greater than 8 days in

b. The dose rate for gaseous effluents sha31 be determined to be iodine-131, iodine-133, within the limits in tri ti um, and for all radionuclides in accordance with the particulate form with methods and procedures half lives greater of the ODCM by ottaining s shall be representative samples than 8 day /yr to and performing analyses

<1500 mrem in accordance with the any organ.

sampling and analysis When the dose rates exceed program specified in the above limi ts, without Table 4.8.2.

delay, decrease the release rate to comply with the limit.

1 20R

C LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIPIME!US 2.

The air dose in areas at 2.

Cumulative dose contributions and beyond the SITE for noble gases shall be BOUNDARY (see Figure determined in accordance 3.8.1) due to noble gases with the methodology and in gaseous effluents released parame:ers in the ODCM from the two reactors at the at least once per month, site shall be limited to the following:

a. During any calendar quarter for gamma radiation

< 10 mrad.

During any calendar quarter for beta rad iation : < 20 mrad.

b. During any calendar year for gamma radiation:

< 20 mrad.

>Euring any< calendar year for beta radiati on: < 40 mrad.

When the calculated air dose from radioactive noble gases in gaseous ef fluents exceeds any of the above limits, prepare and submi t to the Com-mission within 21 working days, pursuant to Speci-fication 6.9.3, a Special Report which identifies the causes for exceeding the limits and defines the corrective actions that have been taken to reduce the releases and proposed corrective actions to be taken to assure that sub-l sequent releases will be I

within the above limits.

Reactor shutdown is not reauired.

3.

The dose to a MEMBER OF 3.

Cumulative dose contributions THE PUBLIC f rom iodine-131, for iodine-131, iodine-133, iodine-133, tritium and triti um, and radionuclides from all radionuclides in particulate form with half in particulate form with

(

lives greater than 8 days

'09 2

l l

P: PS LIMITING CDNDITIONS FOR OPERATION SURVFILLANCE RFOUIPEFEhTF half-lives greater than 8 days in gaseous effluents shall be determined in released from the two accordance with the reactors at the site to methodology and parameters areas at and beyond the in the ODCM at least once

~

SITE BOUNDARY (see Figure per month.

3.8.1) shall be' limited to the following:

o

a. During any calendar quarter: 1 15 mrem.

b. During any calendar year: 1 30 mrem.

When the calculated dose from the rel ease of iodine-131, iodine-133, tritium and

- radionuclides in particulate form, with half-lives greater

'than 8 days in gaseous ef fluents exceeds any of the above limits, prepare and submit to the Commission within 21 working days, pursuant to Speci fication 6.9.3, a Special Report Which identi fies the causes for exceeding the limits and defines the corrective actions that have been taken and proposed corrective actions to assure that sub-sequent releases will be within the above limits.

Reactor shutdown is not reoui r ed.

N During release of gaseous wastes the following con-4a. The reactor building ditions shall be met to exhaust vent and main avoid exceeding the stack noble cas radiation i

limits specified in monitors aball be cali-3.8.C.1:

brated every 17 months with

a. The main off gas stack a known radioactive source minimum dilution flow of positioned in a reproducible 10,000 cfm shall be oeometry with respect to maintained.

the sensor, and every

b. One reactor building quarter by means of a functional test. The exhaust vent monitor channel functional test 210 s

1

1 i

PBAPS LIMITING CVNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS and one main stack noble gas monitor shall also demonstrate shall be operable and set control room alarm an that to alarm in accordance nuncia tion occurs if any of with the methodology the following conditions existe and parameters in the 1.

Instrument indicates ODCM. From and af ter the measured levels date that both reactor sbove the alarm building exhaust vent setpoint.

monitors or both main 2.

Instrument indicates stack noble gas monitors a downscale failure.

)

are made or found to be Additionally, an instrument i

inoperable for any reason, check shall be performed effluent releases via every day.

their respective pathway 4b. The reactor building may continue provided at exhaust vent and the least two independent main stack flow rate grab samples are taken monitors shall be at least once per 8 hrs.

calibrated every 17 and these samples are months. Additionally, an analyzed for gross instrument check shall activity within 24 be performed every day, hours, and at least two 4c. The reactor building l

j technically qualified exhaust vent and the main members of the facility stack iodine and particulate staff independently sampler flow rate monitors verify the release shall be calibrated every rate calculations.

12 months. Additionally,

c. One reactor building an instrument check shall exhaust vent iodine be performed every day.

filter and one main stack iodine filter and one reactor build-ing exhaust vent particulate filter and one main stack particulate filter with their respective flow rate monitors shall be.

operable. From and af ter the date that all iodine filters or all particulate filters for either the i

reactor building exhaust vent monitor or the main stack monitor are made or found to be inoperable for any reason, effluent releases via their respective pathway may 211

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RFOUIRFMFNTS continue provided samples are continuously collected with auxiliary sampling equipment for periods on the order of 7 days and analyzed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the end of the i

sampling period.

d. One reactor building exhaust vent flow rate monitor and one main stack flow rate monitor shall be operable and set to alarm in accordance with the methodology and parameters in the ODCM.

From and after the date that both reactor building exhaust vent flow rate monitors or both main stack flow rate monitors are made or found to be inoperabl e for any reason, effluent releases via their respective pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

e. With less than the minimum number of radioactive gas-eous effluent monitoring instrumentation channels OPERABLE exert best efforts to return the instruments to OPERABLE status within 30 days and if unsuccessful explain in the next Semi-annual Radioactive Effluent Release Peport why the in-operability was not cor-rectedin a timely manner.

5.

Gaseous effluents shall be processed through Sa. Doses due to gaseous the appropriate gaseous ef f3 uent releases to waste treatment system areas at and beyond as described below the SITE BOUFDAPY prior to discharoe shall be projected at least once per 217

PBAPS

. LIMITING CONDITIONS FOR OPERATION SURVEILT.ANCE REOUIREMFt?TF i

a. Gases from the Steam i

I Jet Air Ejector Dis-month in accordance i

charge shall be with the methodology processed through the and parametere in the ODCM.

recombiner, holdup pipe, off gas filter, and off-gas stack.

5b. The appropriate gaseous radioactive waste system

b. Gases from the Mechanical equipment as described Vacuum Pump and Gland Steam in Specification 3.8.C.5 Exhauster discharge shall be demonstrated shall be processed operable every quarter, through the off gas unless utilized to stack.

process gaseous waste during the previous

c. Reactor, turbine, 13 weeks, by analyzing radwaste, and recombiner

.the gaseous waste building atmospheres processed through the shall be processed appropriate eculpment through permanently to determine that it or temporarily installed meets the requirements of equipment in the appropriate Speci fication 3.8.C.I.

tuilding ventilation system and the Reactor Building Sc. An air sample sha3 3 be Ventilation Fxhaust Stack, obtained and analyzed with the axception of the from all building arear following unmonitored with an unmonitored exhausts:

exhaust once per month, 1.

Recirculation M-G Set and Res.ctor Building Cooling Water equipment rooms.

2.

Control room utility and toilet rooms.

3.

Cable spread rocm.

4.

Emergency switchgear rooms.

5.

125/250 VDC Battery i

rooms and the 250 VDC Battery rocms.

6.

Administration Building maintenance decontam-ination area.

With gaseous waste being discharged without treatment as required above, prepare and submit to the Commission within 21 working days 21 3 l

l l

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS pursuant to Specification 6.9.3 a Special Report which

(

includes the following informations

a. Explanation of why gaseous radwaste was being dis-charged without treatment, identification of any inoperable equipment or subsystems and the reason for its inoperability.

b. Action taken to restore the inoperable equipment to operable status,

c. Summary description of action taken to prevent a recurrence.

-Reactor shutdown is not required.

6.

The concentration of hydrogen 6a. An instrument check of the downstream of the recombiners operation of the hydrogen shall be limited to less monitors shall be performed than or equal to 24 by once per day.

volume.

a. With the concentration 6b. The hydrogen monitors and of hydrogen downstream associated alarms downstream of the recombiner greater of the recombiner shall than 24 but less than or be calibrated once per equal to 44 by volume, month.

j restore the concentration to within the limit within 6c. Calibration shall include 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

the use of standard gas

b. With the concentration of samples containing a hydrogen downstream of the nominal recombiner greater than 44 1.

It hydrogen, balance by volume, an orderly nitrogen by volume.

reduction of power shall be 2.

44 hydrogen, balance initiated within one hour nitrogen by volume.

l to bring the hydrogen down-stream of the recombiner to less than or equal to 24 by volume.

c. Except as specified in 3.8.C.6.d, two hydrogen l

monitors downstream of the recombiners shall be eperable during power I

operation.

214

PBAPS lit'ITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

d. With the number of P

hyorogen monitors operable one less than required, operation may continue for up to 14 days provided grab samples are taken and analyzed daily. With both hydrogen monitors inoperable operation may continue for up to 14 days provided grab samples are taken and analyzed every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during power operation.

7a. The radioactivity release 7a. The radioactivity release rate of noble gases from the rate of noble gases from Steam Jet Air Ejector dis-the Steam Jet Air Ejector charge as determined by

discharge shall be determined quantitative _ analysis of to be within limits at the identifiable gamma emitters following frequencies by shall not exceed 320,000 performing an isotopic uCi/sec af ter 30 minutes analysis of a representa-decay. With the radio-tive sample of gases taken activity release rate of at the discharge of the noble gases f rom Steam Steam Jet Air Ejector.

Jet Air Ejector discharge exceeding 320,000 uCi/see 1.

At least once per month af ter 30 minutes decay unless the unit has restore the radioactivity been out of service for release rate to within the entire month,

its limit within 72 hrs.

2.

Within 4 hrs. following or be in hot standby an increase, if the within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

of f-gas monitors indicate an increase of greater than 50% in the steady state fission gas release after factoring out increases due to power changes.

7b. One Steam Jet Air Ejector 7b. The Steam Jet Air Ejector radiatien monitor shall radiation monitors shall be be operable during opera-calibrated every quarter tion of a main condenser and an instrument check Steam Jet Air Ejector.

shall be performed once Upon loss of both steam per day. Additionally a Jet Air Ejector radiation functional test will be 4

monitors, releases may performed every month. The continue via this pathway channel functional test for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided shall also demonstrate that 115

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS temporary monitore are control room alarm an-used. Otherwise, be in at nunciation occurs if any of least HOT STANDBY within the following conditions 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

exist:

1.

Instrument indicates measured levels above the alarm setpoint.

2.

Instrument indicates a downscale f ailure.

. Purging of the primary containment shall be through the Standby Gas Treatment System whenever primary containment integrity is required as specified in 3.7.A.2.

rimary containment purging via the Reactor Building Ventilation Exhaust System may be performed whenever primary containment integrity is not required as specified in 3.7.A.2.

i

'D 40 CFR 190 4.0.D 40 CFR 190 The dose or dose commit-1.

Cumulative dose contribu-ment to a MEMBER OF THE tions from liquid and gaseous PUBLIC f rom all uranium ef fluents shall be determined fuel cycle sources within in accordance with the 8 kilometers is limited methodology and parameters to < 25 mrem to the total in the ODCM.

body or any organ (except 2.

Cumulative dose contribu-the thyroid which is limited tions from direct radiation to <75 mrem) over the from the reactor units calendar year. With the and from radwaste storage calculated dose from the shall be determined in release of radioactive accordance with the method-materials in liquid or ology and parameters in the gaseous ef flue:its, exceeding ODCM.

twice the limits of speci-fi cati ons 3. 8.B. 2, 3.8.C.2, or 3.8.C.3 calculations shall be made to determine whether the limits have been exceeded.

2.

The calculations should be 216

PBAPS i

l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RFOUIREMENTS made, including direct radiation contributions from the reactor units and from I

outside storage tanks to

(

deteruine whether the limits l

have been exceeded. If such is the case, prepare and submit to the Commission, within 21 working days, pursuant to Specification 6.9.3, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and schedule for achieving conformance

~with the above limits.

This'Special--Report shall include an analysis that estimates the radiation exposure to a MEMBER OF THE PUBLIC, including all ef fluent pathways and direct radiation, including

,the releases covered by this report, for the calendar year. It shall also describe levels of radiation and concen-trations of radioactive material involved and the cause of the exposure levels or concentrations.

If the estimated dose exceeds the above limits and if the release condition resulting in violation of 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with 40 CFR 190. Submittal of the report is con-sidered a timely request and a variance is granted until staff action on the request is complete.

i

-216a-I-

-m m-.

PBAPu LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.E Radiological Environmental 4.8.E Radiological Environmental Monitoring Monitoring 1.

All deviations from the 1.

The radiological environmental sampling schedule for the monitoring samples shall be radiological environmental collected at the locatiens monitoring program, as and analyzed as specified required by 4. 8.E.1, shall in Table 4.8.3.a and the be docusented in the annual ODCM. Deviations are permitted report.

from the required sampling

a. When the radiological schedule if specimens are environmental monitoring unobtainable due to hazardous program is not conducted conditions, seasonal unavaila-as described in the ODCM, bility, malfunction of auto-prepare and submit to matic sampling equipment or the Commission, in the other legitimate reasons. If Annual Radiological equipment malf unction occurs, Environmental Operating an ef fort shall be made Reports, a description.

to complete corrective

- of the reasons for not t action prior to the end i

conducting the program of the next sampling period.

as required and the plans for preventing a.

The concentration of radio-a recurrence.

activity as a result of

b. When the level of plant effluents in an radioactivity as the environmental sampling result of plant ef fluents medium shall be evaluated in an environmental on a quarterly basis sampling medium at against the equation:

one or more of the locations specified concent rati on (1) +

in the ODCM exceeds reporting level (1) the reporting levels of Table 4.8.3.b when concentration (2) +

> 1.0

.. ~

averaged over any reporting level (2) calendar quarter, prepare and submit All radionuclides used to the Commission by in this evaluation shall the closing of the be averaged on a month following the calendar quarterly basis,

and of the affected calendar quarter, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the repoiting level of Table 4.8.3.b to be exceeded. The Special Report shall j

also define the corrective l

l 216a -

-g,-

~,-y

,,.---,,-,,w--,-~y-n,y,+,

-gw,mv--

,g-v--7g.+

--m t.-,--,.--y-

. - *, _ -. -.umv,,..,,.w ry,..,.ew-.,,,,,,.wa.-w+--*9-ewe'--e-'

---*-+--w+m-.+.a*'- - - -

seei

1

'BAPS LhMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS actions to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC l

is less than the calendar year reporting level of l

Table 4.8.3.b. When more than one of the radier nuclides in Table 4.8.3.b sampling medium, this report shall be submitted ifs concentrati on (1)

+

reporting level (1) concentrati on (2) +. 11.0 reporting level (2)

When radionuclides other than those in Table 4.9.3.b are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifica-tions. The report is not required if the measured level of radioactivity was not the result of plant ef fluents; however, l

in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. When milk samples become permanently unavailable from any of the sample locations listed in the ODCM, identify l

locations for obtaining replacement samples and add them to the radiological l

environmental monitoring 1

program within 21 working days. Specific locations from which samples are

-216a.-._

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS unavailable may then be deleted f rom the monitoring program. Identify the cause of the unavailability of samples and identify the new locations for ob-taining replacement samples in the next Radioactive Dose Assessment Report and include in the report revised figures and tables for the ODCM reflecting the new locations.

2.

A land use census shall.be 2a. The land use census shall conducted and shall identify be conducte6 every the location of the nearest 12 months by a door-milk animal in each of the to-door survey by 16 meteorological sectors consulting local within a distance of five agriculture authorities miles.

or by some other appropriate means.

a. When a land use census identifies a new loca-tion which yields a calculated dose or dose commitment greater than the values currently being calculated in Speci fication 3.8.C.3, identify the new loca-tion in the next Radioactive Dose Assessment Report.

b. When a land uce census identifies a location which yields a calculated dose or dose commitment (via the same exposure pathway) at least 20%

greater than a location from which samples are c;rrently being obtained in accordance with Specification 3.8.E.1, add the new location to the radiological environmental monitoring l

program within 21 work-ing days. The indicator sampling location having

-216a.- _ _-

l PBAPS i

LIMITING CONDITIONS FOR OPERATION SU3VEILLANCE REQUIREMENTS the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program of ter October 31 of the year in which this land use

]

s census was conducted.

Identify the new location 1

in the next Radio-active Dose Assessment Report and include i

in the report revised 1

figures and tables for the ODCM reflecting the new locations.

3.

Analyses shall be performed 3a. A summary of.the results on radioactive materials obtained as part of the supplied as part of the EPA Interlaboratory Comparison l

Environmental Radioactivity Program shall be included Intercomparison Studies Program, in the Annual Radiological or another Interlaboratory Environmental Operating Comparison Program that has Report pursuant to been approved by the Commission.

Specification 6.9.3.

a. With analyses not being performed as required above report the corrective actions taken to prevent a recurrence in the Annual Radiological Environmental Operating Report.

3.8.F Solid Radioactive Waste 4.8.F Solid Radioactive Waste 1.

The solid radwaste system 1.

The PCP shall be used to shall be used in accordance ensure meeting the burial with a Process Control ground and. shipping re-Program (PCP) to process cuirements prior to shipment wet radioactive wastes to of radioactive wastes from meet shipping and burial the site.

ground reouirements.

a. With the provisions of the Process Control Program not satisfied, suspend shipments of defectively packaged solid radio-active waste from the site. Reactor shutdown is not required.

-216a --

-g

• w

PBAPS TABLE 4.8.1 RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS Sample Lower Limit of Detection Rample Type Sample Freauency Sample Analysis (LLD)(1)(4)(5)

)

1

-7 daste Tank to Each Batch (2)

Quantitative 5 x 10 uCi/ml e released Analysis of Identifiable Gamma Emitters

-6 I-131 1 x 10 uCi/ml

-6

.oporti onal Monthly (3)

Fe-55 1 x 10 uCi/ml mposite of

-5

.ches Tritium 1 x 10 uCi/ml

-7 Gross Alpha 1 x 10 uCi/ml

_a

.oportional Monthly (3)

Sr-89 5 x 10 uCi/ml omposite of

-8 atches Sr-90 5 x 10 uCi/ml

-5

.e Batch Monthly dissolved noble 1 x 10 uCi/ml gases s

Notec 1.

The Sample Lower Limit of Detection is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

The values for the lower l

limit of detection are based on a 95% confidence level.

716b-1

PBAPS 2.

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

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

3.

A composite sample is one in which the quantity of the sample is proportional to the quantity of liquid waste discharged and in which the method of sampling results in a sample representative of the liquids released.

4.

The principal gamma emitters for which the minimum detectable level specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-6 0, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, 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.

Nuclides which are below the sample detectable limit for the analyses should not be reported as being present at the sample detectable limit level.

When unusual circumstances result in sample detectable limits higher than required, the reasons shall be documented in the Semi-

~

Annual E f fluent Report.

The values listed are believed to be attainable.

5.

Certain mixtures of radionuclides may cause interference in the measurement of individual radionuclides at their detectable limit especially if other radionuclides are at much higher concentrations.

Under these circumstances use of known ratios of radionuclides will be appropriate to calculate the levels of such radionuclides.

I t

e 216b-2

PBAPS l

TABLE 4.8.2 RADIOACTIVE GAFFOUS YASTE SAPTLIFG AFD AMAI.YSIS FTOP EAIN OFF-CAS STACK AFD FFACTOR BUII DIt'G VEF7 EXPAUF7 STACK Sample Lower Limit of Sample Type Sample Frecuency Sample Analysis Detecti on (LLD)(3)(4)

-4 Grab Sample Monthly (2)

Quantitative 1 x 10 uCi/cc(3)

Analysis of Identi fiabl e Gamma Er.itters

-6 Grab Sanr3e Ouarterly Tritium 1 x 10 uCi/cc

-12 Charcoal Yeekly(3) 7-131 1y 10 uCi/ce(?)

Fil ters

-10 Particulate reekly(3)

Cuantitative 1 x 10 uCi/cc(?)

Fi3ters Analyris of I dent i fiabl e Gamma Fmitterr

-12 I-131 1 x 10 uCi/ce(?)

-31 Pa rti culat e Monthly Gross Alpha 1 x 10 uCi/cc Filters (composite of veekly filters)

-11 Particulate Fonthly Fr-PP 1 x 30 uCi/cc Filters

-11 (ccmposi te of Sr-90 1 x 10 uCi/cc weekly filters )

Noble Gas Continuously Nobl e Gas

-?

Grost $ orjo 1 x 10 uCi/cc Monitor j

(Main Stack)

NoF3e cas Ccntinuous3y Fob 3e Cas

-F Ponitor Crossjg orjo 3 x 3n uCi/cc (Poof Vents) 216c-1

n e

PBAPS Notes 1.

The Sample Lower Limit of Detection is defined as an _a_ priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

The values for the lower limit of detection are based on a 95% confidence level.

2.

Sampling and analysis shall be performed following shutdown, startup or e thermal power change exceeding 15 percent of rated thermal power within one hour from a steady state ccndition unless (1) analysis shows that the dose equivalent I-131 concentration in the primary coolant has not increased more than a factor of 3, and (2) the noble gas activity monitor shows that effluent activity has not -increased by more than a factor of 3.

l 3.

Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> af ter changing.

Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 3 days following each shutdown, startup or thermal; power change exceeding 15 percent of rated thermal power in one hour and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing.

When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLD may be increased by a factor of 10.

This requirement does not apply if (1) analysis has shown that the dose equivalent I-131 concentration in the primary coolant has not increased more than a f actor of 3, and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

4.

Certain mixtures of radionuclides may cause interforence in the measurement of individual radionuclides at their detectable limit especially if other radionuclides are at much higher concentra tions.

Under these circumstances use of known ratios of radionuclides will be appropriate to calculate the levels of such radi onuclides.

Nuclides which are below the sample detectable limit for the analyses should not be reported as being.present at the sample detectable limit level.

216c-2

r J

7

jyf m

J

~

TABIE 4.B.3.a RADIQU)CIC% CNIBOtMD?TAL KNI'!ORING PIOG%M Number of Samples Exposure Pathway and Sampling and

'Iype and Frecuency and/or sample Sample tocation(a)

Collection Frecuencv(b,c) of Analysis e

1. DIIGCT RADIATION At least 40 routine monitoring At least monthly Gama dose at least stations either with tm or more monthly.

dosimeters or with one instrument or for measering and recording dose or rate continuously to be placed as At least cuarterly fo11cus: 1) an inner ring of sta-Gama dose at least ticns in the general area of the quarterly.

SITE BOUNIAW and an outer ring in the 3 to 6 trile range from the site. A station is in each sector of each ring except as dictated by local e,eography. The balance of the stations are in special interest areas such as population centers, nearby residences, schools and in areas to serve as control stations.

2. AIRBORE Radioicdine f.

Samples from 5 locations:

Continuous sampler Radiolo31ne Cannister_a_

Particulates operation with sample W 1 analysis weekly.

a. 3 samples from close to the collection at least SITE BOUNDARY locations (in weekly or required different sectors) of the by dust loading, Particulate Samler:(f) i highest calculated annual whichever is more Gross beta radio-average groundlevel D/Q.

frequent.

activity analysis follcuing filter

b. 1 sample from the vicinity changer Gama isotopic i

of a conronity having the analysis of ocrnposite highest calculated annual (by location) quarterly.

average grourdlevel D/Q.

c. 1 sample from a control location unlikely to be affected by the plant.

-216d t.

t 0

/

d-4' i

• TAeLE 4.e.3.a (Continued)

RADIOIIGICAL DNIRCrBCMJ, SmI'IURING PKCRAM Msnber of Samoles Exposure Pathway and Samo11rq and Type and Frecaency

__ and/or Samole l

_Samo e Incation(a)

Collection Frecuency(b,c) of Analysis

3. WATERBORNE b

a. Surface

a. I sa vle upstream Ccnposite sample Gama isotopic (d) analysis b.1 sample &wnstream over 1 month renthly. Co w ite for parlod.

tritium analysis at least quarterly.

b

b. Drinking

a. I sample of each of 1~ to 3 Cmposite sample Ocmposite for gross of the nearest water supplies over monthly beta and gama(d) that could be affected by its corposite isotopic analyses discharge.

period.

monthly. Ccs,posite for

b. I sample fmn a control tritium analysis at location least cuarterly.

c. Sediment from i sample frca dcunstream area S ai-annually Garrna isotopic}d)

Shoreline with existing or potentiti recreational value.

analysis sci-annually.

4. INGISFION

a. Milk

a. Samples from milking ' animals Semi-renthly when Cama isotopic (d) analysis in 3 locations within 3 miles animals are on or 134, 137 Cs by distance having the highest pasture, nonthly chemical separation i

dose potential.

at other times.

quarterly. I-131

b. I sample from milking animals analysis of each sample.

at a control location (unlikely to be affected by the plant).

~

s

b. Fish a.1 sample of tach ccruercially Cample in season, oc Ca rna isotopic (d) analysis and recreationally important semi-annually if they on edible portions.

e species in vicinity of dis-are not seasonal charge roir.t when available, b.1 sanple of sane species in areas not influenced by plant dischargt when available.

i

-2166 s 4

4 4,

S y

90 s

4

/

e

~

TABLE 4.8.3.a (Continued)

RADIOII)GICAL DNIHotMDirAL POIMORItC PR1GPAM nrter of Samp14s Exposure Pathway and Sa mling and Type and Fregaency and/or Sample Sample Iccation(a)

Collection Frecuency(b,c) of Analysis

c. Focxl Products

a. Sanples of 3 different kinds Manthly when Cerna isotopic (d) and of broad leaf vegetation grown available if milk I-131 analysis.

nearest offsite garden of sampling is highest annual average ground-not perforined.

level D/0 if milk sampling is not performed.

b. I senple of each of the similar Manthly when Gama isotopic (d) and broad leaf vegetation grown available if milk I-131 analysis.

15-30 km distant in the least sampling is prevalent wind direction if not performed.

milk sampling is not performed.

l N__

a) Fixed sample locations are shown in the Offsite Dose Calculation Manua', Table VII.A.1 and Figures VII.A.1, VII.A.2, and VII. A.3.

At times it may not be possible or practicabTe to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternate media and locations may be chosen for the particular pathway in gaestion.

Such necessary deviations are reported in the Annual Environmental Radiolog! cal Operating

Report, b) Co,posite samples shall be collected by collecting an aliquot at intervals not exceeding two hours.

c) Sample collection fregaencies are defined as follcus Weekly:

7 calendar days + two days Bi-weekly:

14 calendar days'~1 three days Monthly a calendar month + eight days Daarterly:

a three month calendar period i ten days Semi-annually: a six month calendar period i twenty days l

-2166 i I

l l'

1 i

i 1

1 l

.__.___.m______

.._.___.m..

j l

i i

TABIE 4.8.3 (Continued)

RADIO 1DGICAL DNIFDNFNTAL KNIl0 RING P10GRM4_

]t d) Garuna isotopic ar.alysis means the identification and auantification of gerna-enitting l

r;Alonuclides that may be attributable to the effluents fram the facility.

e) Each phosphor is considered one thermoluminescent dosimeter.

i f) If the gross beta activity in air attributable to plant operation is greater than ten times the yearly mean of control samples, gama isotopic analysis shall be performed on the irdividual j

suuples.

i i

1 l

1 0

1 f

1 4

1

- -2166.

a l

)

--m-e --,

s.

,. ~

IRONMENTAL SAMPLES REPORTING LEVELS FO.. J10 ACTIVITY a. u

,-. m i..,

Reporting Levels Water Airborne Particulate Fish Milk Food Products Analysis (oCi/1) or Gases (pci/m3)

(pci/kg. wet)

(pci/1)

(pCi/kg. wet)

H-3 20,0C0*

Mn-54 1,000 30,000 Fe-59 400 10,000 30,000 Co-58 1,000 Co-60 300 10,000 w

Zn-65 300 20,000 e

1 7

Zr-Nb-95 400 Y

0.9 3

100 I-131 l

Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 300 Ba-La-140 200 For drinking water samples.

This is 40 CFP Part 141 value.

r

/

N

(

)

o.

's N 0

400 800 FT.

s' s

N II LDG.

[.,PERIUfd' j

CisRD 1 ROOF VENTS 305' G9 s

h l

9 h

LIQUID 0 ISCH ARGE f)

STRUCTURE l

'g 8403150143-01

\\

PH:L4.DELPHI A ELECTRIC C00 P.B. APS g

UNITS 2 & 3

(

g

\\

EASE 00S AND LIQUID

\\

E. CiUENT RELE ASE POINTS l

......I

\\

TABLE 4.8.3.c MAXIMUM VALUES FOR MINIMUM DETECTABLI

~

LEVELS OF ACTIVITY (MDL) a Food Airborne Fish Products Sediment Water Particulate (pCi/kg, Milk (pCi/kg, (pCi/kg,

. Analysis (pci/1)

(pci/m3) wet)

(pci/1) wet) dry) gross 2.5

.006 beta 3H 1200 54Mn 9

80 59Fe 18 160 38,60Co 9

80 652n 18 160 95Zr-Nb 9

131I

.04 0.6 36 134,137Cs 9,11

.04 90 10 40 100 140Ba-La 9

9 l

1 TABLE NOTATION a - Analyses shall be performed in such a manner that the stated MDLs will be achieved under routine conditions at a 954 confidence level.

Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides or other uncontrollable circumstances make these MDLs unachievable.

(

-216d-6 1

i I

PBAPS 3.8.A & 4. 8. A B ASFS General It is expected that releases of radioactive material in af fluents will be kept at small fractions of the limits specified in Section 20.106 of 10 CFR, Part 20.

At the same time, the licensee is pernitted the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power even under i

unusual operating conditions which may temporarily result in releases higher than such small f ractions, but still within the limits specified in Section 20.176 of 10 CFR, Part 20.

It is expected that in using this operational flexibility under unusual operating conditions the licensee will exert his best efforts to keep levels of' radioactive material in ef fluents as low as practicable.

3. 8.B & 4. 8. B B ASES Concentrati on This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to areas at and beyond the SITE BOUNDARY will be within the concentration levels specified in 10 CFR, Part 20, Appendix B, Table II, Column 2.

This instantaneous limitation provides additional assurance that the levels of radioactive materials in bodies of water in areas at or beyond the SITE BsUNDARY will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR, Part50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR, Part 20.106(e) to the population.

The concentration limit f or noole gases is based upcn the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an eauivalent concentration in I

water using the international Commission on Radiological Protection (ICRP) Publication 2.

l Dose This specification is provided to implement the reauirements of Sections II. A, III.A and IV.A of Appendix I, 10 CFR Part 50.

The Limiting condition for Operation implements the guidance set forth in Section II.A of Appendix I and provides the reouired forth in operating flexibility to implement the guides set Section IV.A of Appendix I to assure that the releases of

-216f _

1 PBAPS l

~

radioactive material in liouid ef fluenta will be kept "as low as reasonably achievable".

The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that I

conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

The equations described in the Off site Dose Calculation Manual for calculating the doses due to the actual release rates of radioactive materials in liquid ef fluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man f rom Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.113,

" Estimating Aquatic Dispersion of Ef fluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", April 1977.

This specification applies to the release of liquid ef fluents f rom the site.

Instrumentation The radioactive liauid ef fluent instrumentation is provided to l

monitor and control, as applicable, the releases of radioactive materials in liould ef fluents during actual or potential release of liquid effluents.

The operability and use of this instrumentation is consistent with the reouirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR, Part 50.

I System Operation The operability of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid ef fluents require treatment prior to release to the environment.

The requirement that the appropriate portions of this system be used when specified provides assurance that the releases to radioactive materials in liquid ef fluents will be kept "as low as i

reasonably achievable".

This specification implements the j

requirements of 10 CFR, Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR, Part 50 and design objective Section II.D of Appendix I to 10 CFR, Part 50.

The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as L suitable fraction of the guide set forth in Section II.A of Appendix I, 10 CFR, Part 50, for liquid ef fluents.

i

-216f l i

l l

PBAPS 3.8.C & 4.8.C BASE _S l

l Dos e 1

l 6

This specification is provided to ensure that the dose from radioactive materials in gaseous ef fluents at and beyond the SITE BOUNMRY will be within the annual dose limits of 10 CFR Part 20.

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 assurance that radioactive material discharged in gaseous ef fluents will not result in the exposure of a MEMBER OF THE PUBLIC, either within or outside the SITE BOUN M RY, to annual average concentrations exceeding *the 1imits specified in Appendix B, Table II' of 10 CFR Part 20.106(b)

For MEMBERS OF. THE PUBLIC who may at times be within the EITE

,BOUNMRY, the occupancy will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above 'that for the SITE BOUNMRY.

Examples of calculations for such MEMBERS OF THE PUBLIC with the appropriate occupancy factors 1

are given in the ODO4 The specified limits restrict, at all times, the gamma and beta dose rates above background to a MEMBER OF THE PUBLIC, at or beyond the SITE BOUNMRY to < 500 mrem / year to the total body or to 13000 arem/ year to the sWin.

These dose rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to i 1500 mrem / year.

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 Limiting Condition for Operation implements the guidance set forth in Section II.B of Appendix I and provides the required operating flexibility to implement the guideo set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as reasonably achievable".

The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance l

with the guidances of Appendix I be shown by calculational procedures based en models and data such that the actual exposure of a MEMBER OF THE PUBLIC through the appropriate pathways is unlikely to be substantially underestimated.

The dose t

calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology prc,vided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from i

Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I",

Revision 1,

-216f -

PBAPS October 1977 and the atmospheric dispersion model submitted on September 30, 1976, in a report titled: "Information Requested in to letter f rom George Lear to E. G. Bauer dated February 17, 1976".

The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditionr.

l Dose - Iodine-131, Tritium and Radionuclides in Particulate Form lais spleification is provided to implement the requirements of t ecti ons II.C, III.A and IV.A of Appendix I,10 CFR Part 50.

The Limiting Conditia, for Operation implements the guidance set forth in Section II.C of Appendix I and provides the required operating flexibility to implement the guides set forth in

'ction IV.A of Appendix I to assure that the releases of dioactive materials in gaseous ef fluents will be kept "as low reasonably achievable".

The ODCM calculational methods aified in the Surveillance Requirements implement the quirements in Section III.A of Appendix I that conformance with ae guides of Appendix I be shown by calculational procedures ased on models and data such that the actual exposure of a EMBER OF THE PUBLIC through appropriate pathways is unlikely to i

• substantially underestimated.

The ODCM calculational methods 3r calculating the doses due to the actual release rates of the abject materials are consistent with the methodology provided in egulatory Guide 1.109, " Calculation of Annual Doses to Man from outine Releases of Reactor Effluents for the Purpose of valuating Compliance with 10 CFR Part 50, Appendix I," Revision

-1, October 1977 and the Atmospheric Dispersion Model submitted on ieptember 30, 1976 in a report titled: "Information Requested in to letter f rom George Lear to E. G. Bauer dated February 17, 1976".

These equations also provide for determining the actual doses based upon the historical average atmospheric onditions.

The release rate specifications for iodine-131, itium, and radionuclides in particulate form with half lives eater than 8 days are dependent on the existing radionuclide

, athways to man in the areas at and beyond the SITE BOUNDARY.

The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radi onuclides, 2) deposition of radionuclides ento green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals grare with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

-216f PBAPS Instr umentati on The radioactive gaseous ef fluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous ef fluents during actual or potential releases of gaseous effluents.

The operability and use of l

instrumentation is consistent with the requirements of General Design criteria 60, 63 and 64 of Appendix A to 10 CFR, Part 50.

System Operation The operability of the gaseous radwaste treatment system ensures that this system will be available for use whenever gaseous ef fluents require treatment prior to release to the environment.

The requirement that appropriate portions of this system be used when specified provides reasonable assurance that the releases of

-radioactive materials in gaseous ef fluents will te kept "as low

.as is reasonably achievable".

This specification implements the requirements of 10 CFR, Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR, Part 50 and design objective Section II.D of Appendix I to 10 CFR, Part 50.

The specified limits governing the use of appropriate portions of the gaseous radwaste treatment system were specified as a suitable fraction of the guidar_ce set forth in Sections II.B and II.C of Appendix I, 10 CFR, Part 50, for gaseous ef fluents.

Main Condenser Restricting the gross radioactivity release rate of noble gases f rom the main condenser provides reasonable assurance that the total body exposure to an individual at the SITE BOUNDARY will not exceed a small fraction of the limits of 10 CFR, Part 100 in the event this ef fluent is inadvertently discharged directly to the envi ronment without treatment.

This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR, Part 50.

Hydrocen Gas Mixture This specification is provided to ensure that the concentration of potentially explosive hydrogen gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen.

Maintaining the concentration of hydrogen below its flammability limits provides assurance that the releases of radioactive materials will be controlled in

-216f - - -

l PBAPP.

conformance with the requirements o' General Design Criterion 60 of Appendix A to 10 CFR Part 50.

i Lontainment Purge l

Specification 3.8.C.8 requires that the primary containment atsosphere receive treatment for the removal of gaseous iodine ano particulates prior to release to provide reasonable assurance that purging operations will not result in exceeding the annual dose limits of 10 CFR Part 20 for areas at or beyond the SITE BOUN DARY.

Total Dose This specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 30.

4 This specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive ef fluents exceed twice the design objective doses of Appendix I.

It is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CPR Part 190 if the individual reactors remain within the reporting requirement level.

The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits.

For the purposes cf the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 kilometers (km) must be considered.

If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions i

resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part l

190.11 and 10 CFR Part 20.405c, is considered to be a timely reauest and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed.

The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1 and 3.11.2.

An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

-216f -.

PBAPS 3.8.E & 4.8.E BASES Monitoring Program The radiological environmental monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those l

radionuclides, which lead to the highest potential radiation I

exposures of MEMBERS OF THE PUBLIC resulting from the two reactors at the site.

This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and supplements the radiological ef fluent 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 ef fluent measurements and the modeling of the environmental exposure pathways.

The reauired detection capabilities for environmental sample analyses are tabulated in terms of the Minimum Detectable Level (MDL).

The MDL's required by Table 4.8.1 and 4.8.2 of the specifications are considered optimum for routine environmental l

measurements in industrial laboratories.

The monitoring program was developed utiliring the experience of the first seven years of commercial operation.

PTogram changes may be initiated based on the additional operational experience.

Land Use Census This specification is provided to ensure that significant changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this censue.

This census satisfies the recuirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.

Interlaboratory Comparison Program The requirement for participation in an Interlaboratory Comparison Program ia provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices, are performed as part of the quality assurance program for environmental monitoring, in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

-216f -

__.-__________J

PBAPS 6.5.1.6 Continued h.

Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Plan, to the Chairman of the operation and Saf ety Review Committee.

i.

Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Plan, to the Chairman of the Operation and Saf ety Review Committee.

j.

Review of every unplanned release reportable under 6.9.2.b.( 5) of radioactive material to the envi rons; evaluate the event; specify remedial action to prevent recurrence; 'and document the

~ event description, evaluttion, and corrective action and the disposition of the corrective action in the plant records.

Authority 6.5.1.7 The Plant Operation Review Committee shall:

a.

Recommend to the Station Superintendent written approval or disapproval of items considered under 6.5.1.6(a) through (d ) above.

b.

Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above cc.antitutes an unreviewed safety question, as defined in 10 CFR 50.59.

c.

Provide immediate written notification to the Superintendent, Generation Division-Nuclear or, in his absence, the Superintendent, Generation Division-Fossil-Hydro, and the Operation and Safety Review Committee of disagreement between the PORC and the Station Superintendent; however, the Station Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

-248-1

_ _ ~...., _..

4 1

PBAPS Records 6.5.1.8 The Plant Operation Review Committee shall maintain written minutes of each meeting and copies shall be 1

provided to the Superintendent, Generation Division-Nuclear and Chairman of the Operation and Safety Review Committee.

l

-248a-

PRAPS 6.5.2.8 Continued The Facility Emergency Plan and implementing e.

procedures at 1:ast once per two years.

f.

The Focility Security Plan and implementing procedures at least once per two years.

The Offsite Dose Calculation Manual and g.

implementing procedures at least once per two years.

h.

The performance of activities reauired by the Quality Assurance Program regarding the radiological monitoring program to meet the provisions of Regulatory Guide 4.1, Revision 1, April 1975, at least once per calendar year.

i.

Any other area of facility operation considered appropriate by the OSR Committee or the Vice President, Electric Production.

Authority 6.5.2.9 The OSR Committee shall report to and advise the Vice President, Electric Production on those areas of responsibility specified to Section 6.5.2.7 and 6.5.2.8.

Records v

6.5.2.10 Records of OSR Committee activities shall be prepared, approved, and distributed as indicated below:

Minutes of each OSR Committee meeting shall be a.

prepared, approved and forwarded to the Vice President, Electric Production within 14 days following each meeting.

b.

Reports of review encompassed by Section 6.5.2.7.e,f,g, and h above, shall be prepared,

approved and forwarded to the Vice President, Electric Production within 14 days following l

completion of the review.

-252-

PBAPS c.

Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice President, Electric Production and to the management positions responsible for the areas audited within 30 days after complation of the audit.

I l

l

-252a-

PBAPS 6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PORC and approved by the Station Superintendent or his designated alternate per Specification 6.1.1 prior to l

implementation and periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made, provided:

The intent of the original procedure is not a.

al t ered,

b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit af f ect ed,

The change is documented, reviewed by the PORC and c.

approved by the Station Superintendent within 14 days of implementation.

6.8.4 Written procedures shall be established, implemented and maintained covering the activities of the radiological ef fluent technical specifications as referenced below a.

Of fsite Dose Calculation Manual b.

Quality Assurance Program for the environmental monitoring using the guidance in Regulatory Guide 4.1, Revision 1, April 1975.

6.9 Reporting Requirements In addition to the applicable reporting requirements of l

Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Administrator of the appropriate Regional Of fice unless otherwise noted.

-254-

PBAPS 6.9.1 Routine Reports a.

Startup Report. A summary report of plant startup l

and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the plant.

The report shall address each of the tests identified in.tha FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of enese values with design predictions and specificati ons.

Any corrective actions that were trequired to obtain satisfactory operation shall

=

also be described.

Any additional specific details required in license conditions based on other commitments shall be reported in this report.

Startup reports shall be submitted within 90 days following resumption or commencement of commercial full power operation.

b.

Annual Occupational Exposure Tabulation (1)

A tabulation shall be made on an annual basis of the number of station utility and othe. personnel (including contractors) receiving exposcres greater than 100 arem/yr and their associated man-rem exposure according to work and job fur.mtion, (2) e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, sh ecial maintenance (describe maintenance), waste processing, and refueling.

This tabulation sha:1 be submitted for the previous calendar year prior to March 1 of each year.

The dose assignment to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.

Small exposure tots 11ing less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole j

body dose received from external sources shall be assigned to specific major job functions.

(1)

A single submittal may be made for a multiple unit

(

station.

(2)

This tabulation supplements the requirements of 10 CFR 20.407.

-254a-

l

(

PBAPS 6.9.2 Continued I

(7)

Conditions arising from natural or man-made events that, as a direct result of the event require plant shutdown, operation of safety systems, or other protective measures required by technical speci fications.

(8)

Errors discoverso in the trcnsient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.

(9)

Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report or technical specificacions that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

Note: This item is intended to provide for reporting of potentially generic problems.

(10) occurrence of an unusual or important radiological event that has potential environmental impact from unit operation, or that has high public interest concerning environmental impact from unit operation.

b.

Thirty Day Written Reports.

The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional Office l

within thirty days of occurrence of the event.

The l

written report shall include, as a minimum, a completed l

copy of a licensee event report form.

Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative l

material to provide complete explanation of the circumstances surrounding the event.

-256-

PBAPS (1)

Reactor protection system or engineered safety f eature instrument settings which are found to be l

less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of l

affected systems.

(2)

Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.

Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurations as described in items 2.b(1) and 2.b(2) need not be reported except where test results themselves reveal a degraded mode as described above.

(3)

Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.

(4)

Abnormal degradation of systems other than those specified in item 2.a(3) above designed to contain radioactive material resulting f rom the fission process.

l Note: Sealed sources or calibration sources are not included under this item.

Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item.

(5)

An unplanned offsite release of 1) more than 1 curie of radioactive material in liquid effluents,

2) more than 150 curies of noble gas from a roof vent or 15,000 curies of noble gas from the stack in gaseous ef fluents, or 3) more than 0.05 curies of radioiodine from a roof vent or 5 curies of radioiodine from the stack in gaseous effluents.

The report of an unplanned off site release of radioactive material shall include a description of the event and equipment involved, the cause(s) of the unplanned release, the actions taken to

-256a-

~_

PEAPS l

prevent recurrence, and the consequences of the unplanned release.

6.9.3 Unioue Reporting Reauirements Special reports shall be submitted to the Director of the appropriate Regional Of fice within the time period

]

specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the reouirements of the applicable reference specification a.

Loss of shutdown margin, Specification 3.3.A and 4.3.A within 14 days of the event.

b.

Reactor vessel inservice inspection, Specification 3.6.G and 4.6.G within 90 days of the completion of the reviews.

c.

Report seismic monitoring instrumentation inoperable for more than 30 days (Specification 3.15B) within the next 10 working days.

Submit a seismic event analysis (Specification 4.15B) within 10 working days of the event.

d.

Primary containment leak rate testing approximately three months af ter the completion of the periodic integrated leak rate test (Type A) required by Specification 4.7.A.2.c.2.

For each periodic test, leakage test results from Type A, B and C tests shall be reported.

B and C tests are local leak rate tests required by Specification 4.7.A.2.f.

The report shall contain an analysis and interpretation of the Type A test results and a summary analysis of periodic Type B and Type C tests that were performed since the last Type A test.

l e.

Calculated dose from release of radioactive I

ef fluents, Specification 3.8.B.2, 3.8.B.4, 3.8.C.2, 3.8.C.3, 3.8.C.6, and 3.8.D.

f.

Sealed source leakage in excess of limits, Specification 3.13.2.

-257-

PBAPS g.

The concentration of radioactivity in excess of the environmental monitoring program reporting levels per Speci fication 3.8.E.lb.

h.

Ef fluent Releases (1)

Annual Radiological Environmental Operating Report Routine radiological environmental operating reports covering the previous calendar shall be submitted prior to May 31 of each year.

The annual radiological environmental operating reports shall include summaries, 1

interpretations, and evaluations of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies with operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

The reports shall also include the results of the annual land une census recuired by Specification 3.8.E.2.

i The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiatior, measurements taken during the period pursuant to the locations specified in Table 4.8.3, as well as summarized and tabulated results of these analyses and measurements in the fermat of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

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

The reports shall miso include or reference f rom previous reports the followings a summary description of the radiological

-258-

j PBAPS environmental monitoring program including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, and i

I measuring eouipment used; at least two maps of all sampling locations keyed to a table giving distances and directions from the midpoint between reactor vents; the results of land use censuues required by Specification 3.8.E.2r the results of the Interlaboratory Comparison Program and discussion of all analyses in which the LLD required by Tables 4.8.1 and 4.8.2 was not achievable.

(2)

Semiannual Radioactive Ef fluent Release Report Routine radioactive effluent release reports covering the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.

The Radioactive Ef fluent Release Reports shall include a summary of the quantities of radioactive lionid and gaseous ef fluents and solid waste released from the site.

The Radioactive L'ffluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped of fsite during the report period a.

Container volume, b.

Total curie quantity (specify whether determined by measurement or estimate),

c.

Principal radionuclides (specify whether determined by measurement or estimate),

d.

Source of waste and prccessing employed (e.g., dewatered spent resin, compact ed dry waste, evaporator bottoms), and

-259-

PBAPS l

e.

Type of container (e.g., LSA, Type A, j

Type B, Large Quantity).

The Radioactive Ef fluent Release Reports shall include a list and description of unplanned releases from the site to areas at and beyond the SITE BOUNDARY of radioactive materials in gaseous and licuid effluents made during the reporting period.

i

(

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the OFFSITE DOSE CALCULATICN MANUAL (ODCE), as well as a listing of new locations for. dose

' calculations and/or environmental monitoring 1cidentified by the land use census pursuant to l-

' Specification 3.8.E.2.

(3)

Radiation Dose Assessment Report The radiation dose assessment reports shall be submitted within 120 days af ter January 1 of each year.

f The Radiation Dose Assessment Report shall include an annual summary of hourly meteorological data collected over the i

previous year.

This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents l

released from the unit or station during the l

previous calendar year.

This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous ef fluent to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY during the report period.

All assumptions used in making these assessments (i.e.,

specific activity, exposure time and location) shall be included in these reports.

The meteorological conditions concurrent with the time of release of radioactive materials j

-259a-l

,_______,______\\

PBAPS in gaseous effluents (as determined by I

sampling frequency and measurement) shall be l

used for determining the gaseous pathway doses.

Approximate methods are acceptable.

l The assessment of radiation doses shall be l

performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

The Radiation Dose Assessment Report shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses f rom primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation.

Guidance for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1; October 1977.

If doses f rom plant effluents do not exceed I

twice the Appendix I limits, a statement to that effect shall constitute a 40 CFR 190 assessment.

~

In lieu of submission with the first half year Radioactive Effluent Release Report, the licensee will retain this summary of requi red meteorological data on site in a file that shall be provided to the NRC upon request.

l i

I

-259b-

PBAPS i

6.10.2 Continued i

d.

Records of radiation exposure for all individuals entering radiation control areas, e.

Records of gaseous and liquid radioactive material released to the environs, f.

Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.

g.

Records of training and qualification for current members of the plant staff.

h.

Records of in-service inspections performed pursuant to these Technical Specifications.

i.

Records of Ouality Assurance activities required by the OA Manual, except as described in 6.10.1 above.

j.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k.

Records of neetings of the PORC and the OSR Committee.

1.

Records for Environmens. ' Qualification which are covered under the provisi.

of paragraph 6.16.

m.

Records of analyses required by the radiological environmental monitoring program that would permit evaluation of the accuracy of the analysis at_4 later date.

This should include procedures effective at specified times and QA records showing that these procedures were followed.

s a

-261-

..--...a

.. -..... - - ~

PBAPS l

6.11 Radiation Protection Program Procedures for personnel radiation protection shall be prepared consistent with requirements of 10 CFR Part 20 l

and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

a.

6.12 Fire Protection Inspections An independent fire protection and loss prevention a.

program inspection shall be performed at least once per 12 months utilizing either qualified off site licensee personnel or an outside fire protection firm.

b.

An inspection of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.

9

+

-261a-

PBAPd 6.17 Offsite Dose Calculation Manual (ODCM) 6.17.1 The ODCM shall describe the methodology and parameters.

to be used in the calculation of of f site doses due to radioactive gaseous and liquid ef fluents.

6.17.2 Licensee initiated changes to the ODCM:

1.

Shall be submitted to the Commission in the Semiannual Radioactive Ef fluent Release Report for the pericx3 in which the change was made.

This subnittal shall contain:

s Suf ficiently detailed information. to totally

-a.

support the rat $cnale for the change without benefit of additional or supplemental informati on.

Information submitted should consist of a package of those pages of the CDCM 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);

b.

A determination that the change will not reduce the accuracy or reliability of dose calculations.

2.

Shall becos.e ef fective upon a date specified and agreed to by the PORC following their review and acceptance. of' the change (s).

6.18 Major Changes to Redioactive Waste Treat' ment Systems 6.18.1 The radioactive waste treatment systems are those systems described in Specifications 3.8.B.3, 3.8.B.d, 3.8.C.4 and 3.8.C.5, whic*n are used to maintain control over radioactive materials in gaseous and licuid ef fluents.

6.18.2 Major changes to the radioactive waste systems shall be made by either of the following methods.

For the purpose of this specification ' major changes' is defined in Specification 6.18.3 below.

-265-

PBAPS A.

Licensee initiated changes:

1)

Licensee initiated changes shall be reported to the Commission as part of the Modification Report reouired by 10 CFR 50.59.

The discussion of each change shall contain:

a.

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

b.

Suf ficient detailed information to totally support the reason for the change without benefit of additional or supplemental information:

c.

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

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

An estimate of the exposure to plant operating personnel as a result of the change; and f.

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

2)

The change shall become effective upon review l

and acceptance by both the PORC and OSR Committee.

-266-

PBAPS B.

Commission initiated changes:

i 1

I 1)

The applicability of the change to the facility shall be determined by the PORC after consideration of the facility design.

I 2)

The licensee shall provide the Commission with written notification of its determination of applicability including any necessary revisions to reflect facility design.

3)

The change shall be reviewed by the OSR Committee at its next regularly scheduled meeting.

4)

The change shall become ef fective on a date proposed by the licensee and confirmed by the Commission.

6.18.3

" Major Changes" to radioactive waste systems shall include the following:

A)

Changes in process equipment, components, structures and af fluent monitoring instrumentation from those described in the Final Safety Analysis Report (FSAR) and evaluated in the staff's Safety Evaluation Report (SER):

B)

Changes in the design of radwaste treatment systems that significantly alter the characteristics and/or quentities of ef fluents

(

released f rom those previously considered in the l

FSAR and SER; C)

Changes in system design which invalidate the accident analysis as described in the SER; and D)

Changes in system design that result in a significant increase in occupational exposure of operating personnel.

-267-

e c

I PROCESS CONTROL PROGRAM i

FOR PEACH BOTTOM ATOMIC POWER STATION j

PROCESS CONTROL PROGRAM FOR PBAPS i

Solid Radioactive Waste The as-built Solid Radioactive Waste l

System is designed to process wet (spent demineralizer resins and spent filter material) and dry solid wastes.

Wet solid wastes are dewatered to preclude free water and packaged in High Integrity containers (HIC).

HICs are designed to contain concentrated radwaste materials generated during the removal of oorrosion and activation products from process systems.

The specific activity of these materials ranges from trace amounts to no greater than 350 uCi/ce.

The predominant radionuclides are Co-60, Cs-137 and En-6 5.

The HICs are molded polyethylene.

This material has outstanding dielectric properties, chemical resistance to solvents, acids,

'and alkalles,. toughness, good barrier properties, high environmental., stress cracking resistance, good cold. impact strength, ultraviolet light stability and radiation resistance.

Testing confirms the HICs ability to withstand vibration, drop, compression, puncture and pressure tests.

Each container receives a variety of quality control checks to confirm its integrity.

HICs are designed to maintain their physical integrity for 10 half-lives of the longest lived significant icotope.

Based on the 30 year half-life of Cs-137, the design HIC lifetime is 300 l

years.

The HICs currently in use at Peach Bottom station are i

utilized for the shipment of solid radwaste to the burial site in l

Barnwell, South Carolina.

These HICs have been accepted for burial of radwaste by the State of South Carolina Department of Health and Environmental Control.

Approved procedures are provided and updated at the station for operation of this system in accordance with design.

l The latest guidance from the NRC staf f redefines the PCP as a document that describes the methods and controls for the i

processing and packaging of solid radioactive waste.

These activities at Peach Bottom are currently controlled by plant procedures to ensure compliance with applicable regulations.

Section 6.8.1 of the current Peach Bottom Technical Specification endorses NRC Reguletory Guide 1.33 which in turn requires solid radwaste procedures.

Section 6.5.1.6 requires the on-site review committee to review these procedures and modifications to the

_.j

, solid radwaste systems.

Modifications are reviewed and reported to the NRC in accordance with 10 CFR 50.59.

The Licensee proposes to describe the PCP in the Peach Bottom FSAR, and revise in accordance with the annual FSAR update regulation.

6 4

s 8

i l

l