NRC Generic Letter 1979-30: Difference between revisions

NRC Generic Letter 1979-030: Transmittal of Rev. 2 of Draft Radiological Effluent Technical Specifications
	ML031320311
Person / Time
Site:	Beaver Valley, Millstone, Salem, Oconee, Palisades, Indian Point, Kewaunee, Point Beach, Prairie Island, Surry, Turkey Point, Haddam Neck, Ginna, Farley, Robinson, San Onofre, Cook, Yankee Rowe, Zion, Trojan
Issue date:	07/18/1979
From:	Gammill W; Office of Nuclear Reactor Regulation
To:	;
References
	NUREG-0472, NUREG-0473 NUDOCS 7908080070, GL-79-030
	Download: ML031320311 (78)
	v • d • e

Latest revision as of 01:55, 24 November 2019

UNITED STATES

v o NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 2055,- July 18, 1979

6L- ?eC'ago ALL PRESSURIZED WATER REACTOR LICENSEES

Gentlemen:

Enclosed is a copy of Revision 2 of the Draft-Radiological Effluent Technical Specifications, (NUREG-0472, PWR; NUREG-0473, BWR). This revision embodies corrections of errors found in previous revisions, revisions in format for consistency with Appendix "A" Standard Technical Specifications, and some rewriting for clarification.

If you have prepared Radiological Effluent Technical Specifications to a previous revision but have not submitted them to NRC, do not convert them to this Revision 2 format. You should submit them in their present format since the information required for Revision 2 is available in previous revisions. Otherwise, Revision 2 should be used to prepare your Technical Specification submittal.

In those cases where submittals of Radiological Effluent Technical Specifications in a previous format are made to NRC, the final specifications, after negotia- tions and reviews have been completed, will be based on Revision 2.

If you have any questions, please contact us.

Sincerely, (cy.6 f William P. Gammill, Acting Assistant Director for Operating Reactor Projects Division of Operating Reactors Enclosure:

NUREG-0472, Revision 2

7 90808007 0

N-

NUREG-0472 REVISION 2 RADIOLOGICAL EFFLUENT TECHNICAL

SPECIFICATIONS FOR PWR'S

JULY 1979 AOO8080 0 87

INDEX

DEFINITIONS

SECTION PAGE

1.0 DEFINITIONS

Channel Calibration .................................... 1-1 Channel Check .................................... 1-1 Channel Functional Test .................................... 1-1 Dose Equivalent I-131......................................:-....... 1-1 Source Check.................................... 1-1 Process Control Program (PCP) .................................... 1-1 Solidification..................................................... 1-7 Offsite Dose Calculation Manual (ODCM) ............................. 1-7 Gaseous Radwaste Treatment System .................................. 1-7 Ventilation Exhaust Treatment System ............................... 1-7 Purge-Purging .................................... 1-7 Venting............................................................ 1-7 NOTE: Add 3/4.3.3.9 and 3/4.3.3.10 with appropriate page numbers to Index section for Monitoring Instrumentation and its Bases; also add 5.1.3 and 5.1.4 to Section 5.0 Index.

PWR-STS-I I

. -

INDEX

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION PAGE

3/4.11 RADIOACTIVE EFFLUENTS

3/4.11.1 LIQUID EFFLUENTS

Concentration............................................ 3/4 11-1 Dose..................................................... 3/4 11-5 Liquid Waste Treatment................................... 3/4 11-6 Liquid Holdup Tanks...................................... 3/4 11-7

3/4 11.2 GASEOUS EFFLUENTS

Dose Rate................................................ 3/4 11-8 Dose-Noble Gases......................................... 3/4 11-12 Dose-Radioiodines, Particulate, and Radionuclides Other than Noble Gases................... 3/4 11-13 Gaseous Waste Treatment.................................. 3/4 11-14 Explosive Gas Mixture.................................... 3/4 11-15 Gas Storage Tanks........................................ 3/4 11-17

3/4 11.3 SOLID RADIOACTIVE WASTE.................................. 3/4 11-18

3/4 11.4 TOTAL DOSE............................................... 3/4 11-20

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING

3/4.12.1 MONITORING PROGRAM...................................... 3/4 12-1

3/4.12.2 LAND USE CENSUS......................................... 3/4 12-10

3/4.12.3 INTERLABORATORY COMPARISON.............................. 3/4 12-11 PWR-STS-I II

INDEX

BASES

SECTION PAGE

3/4.11 RADIOACTIVE EFFLUENTS

3/4.11.1 LIQUID EFFLUENTS ........................................ B 3/4 11-1

3/4.11.2 GASEOUS EFFLUENTS ....................................... B 3/4 11-2

3/4.11.3 SOLID RADIOACTIVE WASTE ................................. B 3/4 11-5

3/4.11.4 TOTAL DOSE .............................................. B 3/4 11-5

3/4.12 RADIOACTIVE ENVIRONMENTAL MONITORING

3/4.12.1 MONITORING PROGRAM ...................................... B 3/4 12-1

3/4.12.2 LAND USE CENSUS ......................................... B 3/4 12-1

3/4.12.3 INTERLABORATRY COMPARISON PROGRAM ....................... B 3/4 12-1 PWR-STS-I III

\ ,-

1.0 DEFINITIONS

CHANNEL CALIBRATION

of the

1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, accuracy to channel output such that it responds with the necessary range and CHANNEL CALIBRA-

known values of the parameter which the channel monitors. The alarm and/or TION shall encompass the entire channel including the sensor and FUNCTIONAL TEST. The CHANNEL

trip functions, and shall include the CHANNEL or total CALIBRATION may be performed by any series of sequential, overlapping channel steps such that the entire channel is calibrated.

CHANNEL CHECK

behavior

1.10 A CHANNEL CHECK shall be the qualitative assessment of channel where during operation by observation. This determination shall include, other indica- possible, comparison of the channel indication and/or status with measuring tions and/or status derived from independent instrumentation channels the same parameter.

CHANNEL FUNCTIONAL TEST

1.11 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY

including alarm and/or trip functions.

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

DOSE EQUIVALENT I-131

1.19 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie/

gram) which alone would produce the same thyroid dose as the quantity and I-132, I-133, I-134 and I-135 actually present.

isotopic mixture of I-131, be those The thyroid dose conversion factors used for this calculation shall for Power listed in Table III of TID-14844, "Calculation of Distance Factors and Test Reactor Sites."

SOURCE CHECK

response

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

PROCESS CONTROL PROGRAM (PCP)

and

1.30 The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, wastes from formulation determination by which SOLIDIFICATION of radioactive liquid systems is assured.

PWR-STS-I 1 -1

i a

1.0 DEFINITIONS (Continued)

SOLIDIFICATION

1.31 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

OFFSITE DOSE CALCULATION MANUAL (ODCM)

1.32 The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of of gaseous and liquid effluent monitoring alarm/trip setpoints.

GASEOUS RADWASTE TREATMENT SYSTEM

1.33 A GASEOUS RADWASTE TREATMENT SYSTEM is 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.

VENTILATION EXHAUST TREATMENT SYSTEM

1.34 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

PURGE - PURGING

1.35 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 required to purify the confinement.

VENTING

1.36 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or not other operating condition, in such a manner that replacement air or gas is provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

PWR- STS- I 1-7

TABLE 1.2 FREQUENCY NOTATION

NOTATION FREQUENCY

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

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

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

P Completed prior to each release.

N.A. Not applicable.

PWR-STS-I 1-8

INSTRUMENTATION

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION

LIMITING CONDITION FOR OPERATION

3.3.3.9 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm/

trip setpoints of these channels shall be determined in accordance with the offsite Dose Calculation Manual (ODCM).

,"'PLICABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or decleare the channel inoperable.

b. With one or more radioactive liquid effluent monitoring instrumenta- tion channels inoperable, take the ACTION shown in Table 3.3-12.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE

CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12.

PWR-STS-I 3/4 3-65

C

TABLE 3.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION

MINIMUM

CHANNELS

INSTRUMENT OPERABLE ACTION

1. GROSS RADIOACTIVITY MONITORS PROVIDING AUTOMATIC

TERMINATION OF RELEASE

Liquid Radwaste Effluent Line (1) 28 a.

Steam Generator Blowdown Effluent Line (1) 29 b.

Turbine Building (Floor Drains) Sumps Effluent Line (1) 30

c.

2. GROSS RADIOACTIVITY MONITORS NOT PROVIDING AUTOMATIC

v ~ TERMINATION OF RELEASE

(1) 30

a. Service Water System Effluent Line cm Component Cooling Water System Effluent Line (1) 30

b.

3. CONTINUOUS COMPOSITE SAMPLERS AND SAMPLER FLOW MONITOR

Steam Generator Blowdown Effluent Line (1) 29 a.

Turbine Building Sumps Effluent Line (1) 30

b.

4. FLOW RATE MEASUREMENT DEVICES

Liquid Radwaste Effluent Line (1) 31 a.

Discharge Canal (1) 31 b.

Steam Generator Blowdown Effluent Lines (1) 31 c.

.

TABLE 3.3-12 (Continued)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMEN rATION

I-

MINIMUM

CHANNELS

INSTRUMENT OPERABLE ACTION

5. RADIOACTIVITY RECORDERS (*)

a. Liquid Radwaste Effluent Line (1). 33 'I-

b. Steam Generator Blowdown Effluent Line (1) 34

6. TANK LEVEL INDICATING DEVICES (for tanks outside plant buildings)

a. (1) 32 b. (1) 32 C. (1) 32 d. (1) 32

(

(xRequired only if alarm/trip set point is based on recorder-controller)

TABLE 3.3-12 (Continued)

TABLE NOTATION

the ACTION 28 - With the number of channels OPERABLE less than required bymay Minimum Channels OPERABLE requirement, effluent releases continue for up to 14 days provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.3, and b. At,.-least two technically qualified members of the Facility Staff independently verify the Release rate calculations and discharge line valving; -

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab at samples are analyzed for gross radioactivity (beta or gamma)

a limit of detection of at least 10 microcuries/gram:

a. At least once per.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcuries/gram DOSE EQUIVALENT 1-131.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 micro- curies/gram DOSE EQUIVALENT 1-131.

ACTION 30 - With the number of channels OPERABLE less than required by the via Minimum Channels OPERABLE requirement, effluent releases that, at this pathway may continue for up to 30 days provided least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed for gross radioactivity (beta or-gamma) at a limit of detection of at least 10 microcuries/ml.

ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves may be used to estimate flow.

ACTION 32 - With the number of channels OPERABLE less than required by the this Minimum Channels OPERABLE requirement, liquid additions to tank may continue for up to 30 days provided the tank liquid level is estimated during all liquid additions to the tank.

3/4 3-68

TABLE 3.3-12 (Continued)

TABLE NOTATION

ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 14 days provided the gross radioactivity level is determined at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

ACTION 34 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided the gross radioactivity level is determined at least once per

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual release.

PWR-STS-I 3/4 3-69

TABLE 4.3-12

,0 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

Ln CHANNEL

CHANNEL SOURCE CHANNEL FUNCTIONAL

INSTRUMENT CHECK CHECK CALIBRATION TEST

1. GROSS BETA OR GAMMA RADIOACTIVITY MONITORS

PROVIDING ALARM AND AUTOMATIC TERMINATION

OF RELEASE (

a. Liquid Radwaste Effluents Line D P R(3) Q(M)

b. Steam Generator Blowdown Effluent Line D M R(3) Q(M)

c. Turbine Building (Floor Drains) Sumps Effluent Line D M R(3) Q(M)

w w

2. GROSS BETA OR GAMMMA RADIOACTIVITY MONITORS

0 PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC

TERMINATION OF RELEASE

a. Service Water System Effluent Line D M R(3) Q(2)

b. Component Cooling Water System Effluent Line D M R(3) Q(2) (

3. CONTINUOUS COMPOSITE SAMPLERS AND SAMPLER

FLOW MONITOR

a. Steam Generator Blowdown Effluent Line D N.A. R Q

b. Turbine Building Sumps Effluent Line 0 N.A. R Q

it

TABLE 4.3-12 (Continued)

-0

Z-

M RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

VI--4 LA

I CHANNEL

CHANNEL SOURCE CHANNEL FUNCTIONAL

CHECK CHECK CALIBRATION TEST

INSTRUMENT

4. FLOW RATE MEASUREMENT DEVICES

D(4) N.A. R Q

a. Liquid Radwaste Effluent Line D(4) N.A. R Q

b. Steam Generator Blowdown Effluent Line D(4) N.A. R Q

C

c. Discharge Canal

5. RADIOACTIVITY RECORDERS

4 a. Steam Generator Blowdown Effluent Line 0 N.A. R Q

w

0 N.A. R Q

b. Liquid Radwaste Effluent Line

6. TANK LEVEL INDICATING DEVICES (for tanks outside the building)

a. D* N.A. R Q

D* N.A. R Q

b.

D* N.A. R Q

C.

D* N.A. R Q

d.

TABLE 4.3-12 (Continued)

TABLE NOTATION

During liquid additions to the tank.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

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

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (Operating plants may substitute previously established calibration procedures for this requirement.)

(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.

PWR- STS- I 3/4 3-72

INSTRUMENTATION

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

LIMITING CONDITION FOR OPERATION

3.3.3.10 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the ODCM.

APPLICABILITY: As shown in Table 3.3-13 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable.

b. With one or more radioactive gaseous effluent monitoring instrumenta- tion channels inoperable, take the ACTION shown in Table 3.3-13.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE

CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-13.

PWR- STS- I 3/4 ,3-73

TABLE 3.3-13 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

4j_

MINIMUM CHANNELS

nrnADI C APPLICABILITY ACTION

INSTRUMENT ur'EMMULL

1. WASTE GAS HOLDUP SYSTEM

35 a. Noble Gas Activity Monitor (1)

41 b. Iodine Sampler (1)

41 c. Particulate Sampler (1)

36 d. Effluent System Flow Rate Measuring Device (1)

36

-0 e. Sampler Flow Rate Measuring Device (1)

4A)

~4 2A. WASTE GAS HOLDUP SYSTEM EXPLOSIVE GAS

MONITORING SYSTEM (for systems designed to withstand the effects of a hydrogen explosion)

39 a. Hydrogen Monitor (1) **

39 b. Hydrogen or Oxygen Monitor (1)

2B. WASTE GAS HOLDUP SYSTEM EXPLOSIVE GAS

MONITORING SYSTEM (for systems not designed to withstand the effects of a hydrogen explosion)

(2) **

40

a. Hydrogen Monitor

(2) 40

b. Hydrogen or Oxygen Monitor

TABLE 3.3-13 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

MINIMUM CHANNELS

OPERABLE APPLICABI LITY ACTION

INSTRUMENT

3. CONDENSER EVACUATION SYSTEM

I

a. Noble Gas Activity Monitor (1) 37 b. Iodine Sampler

- (1) 41

c. Particulate Sampler (1)

41

(

Flow Rate Monitor (1) 36 d.

Sampler Flow Rate Monitor (1) 36 e.

--j C",

4. VENT HEADER SYSTEM

37 a. Noble Gas Activity Monitor (1)

Iodine Sampler (1) 41 b.

Particulate Sampler (1) 41 c.

Flow Rate Monitor (1) 36 d.

Sampler Flow Rate Monitor (1) 36 e.

5. CONTAINMENT PURGE SYSTEM

(1) 38 a. Noble Gas Activity Monitor

(1) 41 b. Iodine Sampler Particulate Sampler (1) 41 c.

TABLE 3.3-13 (Continued)

-o

C RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

(A

MINIMUM CHANNELS

OPERABLE APPLICABILITY ACTION

INSTRUMENT

5. CONTAINMENT PURGE SYSTEM (Continued)

e. Flow Rate Monitor (1) 36 (

(1) 36 f. Sampler Flow Rate Monitor

6. AUXILIARY BUILDING VENTILATION

SYSTEM

37 a. Noble Gas Activity Monitor (1)

4aP. *

(1) 41 c4M b. Iodine Sampler *

(1) 41 c. Particulate Sampler *

(1) 36 d. Flow Rate Monitor *

(1) 36 e. Sampler Flow Rate Monitor

7. FUEL STORAGE AREA VENTILATION SYSTEM

37 a. Noble Gas Activity Monitor (1)

(1) 41 b. Iodine Sampler *

(1) 41 c. Particulate Sampler *

(1) 36 d. Flow Rate Monitor *

(1) 36 e. Sampler Flow Rate Monitor

TABLE 3.3-13 (Continued)

-o RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

-94 (U,

MINIMUM CHANNELS

OPERABLE APPLICABILITY ACTION

INSTRUMENT

8. RADWASTE AREA VENTILATION SYSTEM

(1) 37 a. Noble Gas Activity Monitor

(1) 41 b. Iodine Sampler

41 c. Particulate Sampler (1)

(

(1) 36 d. Flow Rate Monitor

l

(1) 36 e. Sampler Flow Rate Monitor

9. STEAM GENERATOR BLOWDOWN VENT SYSTEM

A

(1) 37 a. Noble Gas Activity Monitor

(1) 41 b. Iodine Sampler

(1) 41 c. Particulate Sampler *

(1) 36 d. Flow Rate Monitor

(1) 36 e. Sampler Flow Rate Monitor

4 TABLE 3.3-13 (Continued)

TABLE NOTATION

At all times.

- During waste gas holdup system operation (treatment for primary system offgases).

ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment for up to 14 days provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, and b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup;

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 38 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of this waste gas holdup system may continue for up to 30 days provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 40 - With the number of channnels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of this system may continue for up to 14 days. With (two) channels inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 41 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.11-2.

PWR-STS-I 3/4 3-78

TABLE 4.3-13 a

RAnTDArTTVF GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

--i (A

CHANNEL MODES IN WHICH

I-

CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE

CHECK CHECK CALIBRATION TEST REQUIRED

INSTRUMENT

1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity Monitor P P R(3) Q(l) *

WI N.A. N.A. N.A.

b. Iodine Sampler *

W N.A. N.A. N.A. *

(

c. Particulate Sampler d. Flow Rate Monitor P N.A. R Q

w e. Sampler Flow Rate Monitor D N.A. R Q

.bi t 2. WASTE GAS HOLDUP SYSTEM EXPLOSIVE

GAS MONITORING SYSTEM

D N/A Q(4) M

a. Hydrogen Monitor **

0 Q(4) M

b. Hydrogen Monitor (alternate) N/A

D Q(5) M

c. Oxygen Monitor N/A

D N.A. Q(5) M ** (j d. Oxygen Monitor (alternate)

TABLE 4.3-13 (Continued)

-I

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

(A~

CHANNEL MODES IN WHICH

CHANNEL FUNCTIONAL SURVEILLANCE

CHANNEL SOURCE REQUIRED

CHECK CHECK CALIBRATION TEST

INSTRUMENT

3. CONDENSER EVACUATION SYSTEM

x a. Noble Gas Activity Monitor D M R(3) Q(2) (

W N.A. N.A. N.A.

b. Iodine Sampler c. Particulate Sampler W N.A. N.A. N.A. *

d. Flow Rate Monitor D N.A. R Q *

la

4b e. Sampler Flow Rate Monitor D N.A. R Q

l co *

CD

4. VENT HEADER SYSTEM

D M R(3) Q(2)

a. Noble Gas Activity Monitor

W N.A. N.A. N.A.

b. Iodine Sampler W N.A. N.A. N.A.

(

c. Particulate Sampler

d. Flow Rate Monitor D N.A. R Q

e. Sampler Flow Rate Monitor D N.A. R Q

TABLE 4.3-13 (Continued)

-o REQUIREMENTS

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE

-I

CHANNEL MODES IN WHICH

CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE

CHECK CALIBRATION TEST REQUIRED

INSTRUMENT CHECK

5. CONTAINMENT PURGE SYSTEM *

D P R(3) Q(l)

a. Noble Gas Activity Monitor *

W N.A. N.A. N.A.

b. Iodine Sampler *

W N.A. N.A. N.A.

c. Particulate Sampler *

d. Flow Rate Monitor D N.A. R Q

'A,

1-~

D N.A. R Q

e. Sampler Flow Rate Monitor CA,

IA

6. AUXILIARY BUILDING VENTILATION SYSTEM

D M R(3) Q(2)

a. Noble Gas Actvity Monitor *

W N.A. N.A. N.A.

b. Iodine Sampler W N.A. N.A. N.A.

c. Particulate Sampler D N.A. R Q *

d. Flow Rate Monitor (

e. Sampler Flow Rate Monitor D N.A. R Q

7. FUEL STORAGE AREA VENTILATION SYSTEM

A

D M R(3) Q(2)

a. Noble Gas Activity Monitor A

W N.A. N.A. N.A.

b. Iodine Sampler *

W N.A. N.A. N.A.

c. Particulate Sampler

TABLE 4.3-13 (Continued)

-vi RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

-I

CHANNEL MODES IN WHICH

CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE

INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED

7. FUEL STORAGE AREA VENTILATION SYSTEM (Continued)

(

d. Flow Rate Monitor e. Sampler Flow Rate Monitor D

D

N.A.

R

Q

I

t

8. RADWASTE AREA VENTILATION SYSTEM

A

4- a. Noble Gas Activity Monitor D M R(3) Q(2)

b. Iodine Sampler W N.A. N.A. N.A

04I

c. Particulate Sampler W N.A. N.A. N.A

A

d. Flow Rate Monitor D N.A. R Q

e. Sampler Flow Rate Monitor D N.A. R Q

(

9. STEAM GENERATOR BLOWDOWN VENT

A

a. Noble Gas Activity Monitor D M R(3) Q(2)

b. Iodine Sampler W N.A. N.A. N.A

A

c. Particulate Sampler W N.A. N.A. N.A.

d. Flow Rate Monitor D N.A. R Q

A

e. Sampler Flow Rate Monitor D N.A. R Q

TABLE 4.3-13 (Continued)

TABLE NOTATION

At all times.

- During waste gas holdup system operation (treatment for primary system offgases).

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

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

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (Operating plants may substitute previously established calibration procedures for this requirement.)

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1. One volume percent oxygen, balance nitrogen, and

2. Four volume percent oxygen, balance nitrogen.

PWR-STS-I 3/4 3-83

I

3/4.11 RADIOACTIVE EFFLUENTS

3/4.11.1 LIQUID EFFLUENTS

CONCENTRATION

LIMITING CONDITION FOR OPERATION

3.11.1.1 The concentration of radioactive material released from the site (see Figure 5.1-4) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved_gr entrained noble gases, the concentration shall be limited to 2 x 10 microcuries/ml total activity.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released from the site exceeding the above limits, immediately restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 4.11-1. The results of pre-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.11.1.1.

4.11.1.1.2 Post-release analyses of samples composited from batch releases shall be performed in accordance with Table 4.11-1. The results of the previous post-release analyses shall be used with the calculational methods in the ODCM to assure that the concentrations at the point of release were maintained within the limits of Specification 3.11.1.1.

4.11.1.1.3 The radioactivity concentration of liquids discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 4.11-1. The results of the analyses shall be used with the calculational methods in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1.

PWR-STS-I 3/4 11-1

TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM

Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD) a Type Frequency Frequency Analysis (,Ci/ml)

A. Batch Waste P P

Release Each Batch Each Batch Principa} Gamma 5xlO

Tanksd Emitters I-131 lxl0 6

-5 P M Dissolved and 1x10

One Batch/M Entrained Gases (Gamma emitters)

1x10 5 P M H-3 Each Batch Compositeb Gross Alpha x10 77 P-32 lx10-6

5x10 8 P Q Sr-89, Sr-90

Each Batch Composite 6 Fe-55 x10-6

5xlO 7 B. Continuous W Principal Gamma Releasese Continuousc Composite Emitters I-131 1x10- 6 M M Dissolved and lxlO 5 Grab Sample Entrained Gases (Gamma Emitters)

lxlO 5 M H-3 Continuous c Compositec 7 Gross Alpha 1x1O0

lxl0 6 P-32

5xlO 8 Q Sr-89, Sr-90

ContinuousCc CompositeC 6 Fe-55 lxlO

PWR-STS-I 3/4 11-2

.

TABLE 4.11-1 (Continued)

TABLE NOTATION

a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 s LLD = E

V

2.22

Y

exp (-AAt)

Where:

LLD is the lower limit of detection as defined above (as picocurie per unit mass or volume),

s is the standard deviation of the background counting rate or of tee counting rate of a blank sample as appropriate (as counts per minute),

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

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

2.22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

The value of s used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples.

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

The background count rate is calculated from the background counts that are determined to be within + one FWHM (Full-Width-at-Half-Maximum)

energy band about the energy of the gamma ray peak used for the quantitative analysis for that radionuclide.

PWR- STS- I 3/4 11-3

TABLE 4.11-1 (Continued)

TABLE NOTATION

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

c. To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be throughly mixed in order for the composite sample to be representative of the effluent release.

d. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, by a method described in the ODCM, to assure representative sampling.

e. A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume of system that has an input flow during the continuous release.

f. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Cs-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 LLD for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations.

PWR- STS- I 3/4 11-4

RADIOACTIVE EFFLUENTS

DOSE

LIMITING CONDITION FOR OPERATION

3.11.1.2 The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site (see Figure 5.1-4) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters, so that the cumulative dose or dose commitment to an individual from such releases during these four calendar quarters is within 3 mrem to the total body and 10 mrem to any organ. (This Special Report shall also include (1) the results of radiological analyses of the drinking water source and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act.*)

b. The provisions of specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days.

Applicable only if drinking water supply is taken from the receiving water body.

PWR-STS- I 3/4 11-5

RADIOACTIVE EFFLUENTS

LIQUID WASTE TREATMENT

LIMITING CONDITION FOR OPERATION

3.11.1.3 The liquid radwaste treatment system shall be OPERABLE. The appro- priate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site (see Figure 5.1-4) when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the liquid radwaste treatment system inoperable for more than

31 days or with radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and

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

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases shall be projected at least once per

31 days, in accordance with the ODCM.

4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE

by operating the liquid radwaste treatment system equipment for at least minutes at least once per 92 days unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 days.

PWR-STS-I 3/4 11-6

RADIOACTIVE EFFLUENTS

LIQUID HOLDUP TANKS*

LIMITING CONDITION FOR OPERATION

3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited to less than or equal to curies, excluding tritium and dissolved or entrained noble gases.

a.

b.

C.

d. Outside temporary tank APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.

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

PWR-STS-I 3/4 11-7

RADIOACTIVE EFFLUENTS

3/4.11.2 GASEOUS EFFLUENTS

DOSE RATE

LIMITING CONDITION FOR OPERATION

3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site (see Figure 5.1-3) shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin, and b. For all radioiodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, immediately decrease the release rate to within the above limit(s).

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM.

4.11.2.1.4 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program, specified in Table 4.11-2.

PWR-STS-I 3/4 11-8

TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM

(A~

-.4 Minimum Lower Limit of Sampling Analysis Type of Detection (hLD)

Gaseous Release Type Frequency Frequency Activity Analysis (zCi/m1)

_P PP

x10 4 A. Waste Gas Storage Each Tank Each Tank Principal Gamma Emittersg Tank Grab Sample c.F*

B.Cnaimn Preb b 9- 4 I

to4 B. Containment Purge Each Purge Each Purge Principal Gamma Emitters x10

Grab -

Sample H-3 lxlO (

C. (List other release Mb,ce Mb Principal Gamma Emittersg lxl0&4 points where gas- Grab lx1 6 eous effluents are Sample H-3 discharged from the facility)

D. All Release Types Continuousf Wd I-131 1xl0 12 as listed in A, B, Charcoal -10

C above. Sample I-133 xl10

Continuousf Wd Principal Gamma Emittersg lx 10 11 Particulate (I-131, Others)

Sample lxlO 11 Continuousf M Gross Alpha Composite Particulate Sample

(

Continuousf Q Sr-89, Sr-90 lxlO 11 Composite Particulate Sample Continuousf Noble Gas Noble Gases lxlO 6 Monitor Gross Beta & Gamma

TABLE 4.11-2 (Continued)

TABLE NOTATION

a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 s LLD =

E -

2.22

Y

exp (-XAt)

Where:

LLD is the lower limit of detection as defined above (as picocurie per unit mass or volume),

s is the standard deviation of the background counting rate or of tbe counting rate of a blank sample as appropriate (as counts per minute),

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

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

2.22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples.

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

The background count rate is calculated from the background counts that are determined to be within + one FWHM (Full-Width-at-Half-Maximum)

energy band about the energy of the gamma ray peak used for the quantitative analysis for that radionuclide.

PWR-STS-I 3/4 11-10

TABLE 4.11-2 (Continued)

TABLE NOTATION

b. Analyses shall also be performed following shutdown, startup, or similar operational occurrence which could alter the mixture of radionuclides.

c. Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.

d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after charging (or after removal from sampler). Sampling and analyses shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup in or thermal power level change exceeding 15% of RATED THERNAL POWER

the cor- one hour. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, responding LLD's may be increased by a factor of 10.

e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area.

f. The ratio of the sample flow rate to the sampled steam flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1,

3.11.2.2 and 3.11.2.3.

g. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuc- lides are to be detected and reported. Other peaks which are measure- able and identifiable, together with the above nuclides, shall also be identified and report. Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD and shall not be reported as being present at the LLD level for that nuclide.

The "less than" values shall not be used in the required dose calcul- tions.

PWR-STS-I 3/4 11-11

RADIOACTIVE EFFLUENTS

DOSE - NOBLE GASES

LIMITING CONDITION FOR OPERATION

3.11.2.2 The air dose due to noble gases released in gaseous effluents from the site (see Figure 5.1-3) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and, b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

(The dose design objectives shall be reduced based on predicted noble gas releases from the turbine building if effluent sampling is not provided. The dose design objectives shall also be reduced based on expected public occupancy of areas, e.g., beaches and visitor centers within the site boundary.)

APPLICABILITY: At all times.

ACTION

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceedigng any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters, so that the cumulative dose during these four calendar quarters is within (10) mrad for gamma radiation and (20) mrad for beta radiation.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the OCM at least once every 31 days.

PWR-STS-I 3/4 11-12

RADIOACTIVE EFFLUENTS

DOSE - RADIOIODINES, RADIOACTIVE MATERIALS IN PARTICULATE FORM, AND RADIONUCLIDES

OTHER THAN NOBLE GASES

LIMITING CONDITION FOR OPERATION

3.11.2.3 The dose to an individual from radioiodines and radioactive materials in particulate form, and radionuclides (other than noble gases) with half-lives

5.1-3)

greater than 8 days in gaseous effluents released from the site (see Figure shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and, b. During any calendar year: Less than or equal to 15 mrem to any organ.

(The dose design objectives shall be reduced based on predicted carbon-14 releases and turbine building releases if effluent sampling is not provided.)

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides (other than noble gases) with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases of radio- iodines and radioactive materials in particulate form, and radio- nuclides (other than nobles gases) with half-lives greater than

8 days in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters, so that the cumulative dose or dose commitment to an individual from such releases during these four calendar quarters is within (15) mrem to any organ.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once every 31 days.

PWR- STS- I 3/4 11-13

RADIOACTIVE EFFLUENTS

GASEOUS RADWASTE TREATMENT

LIMITING CONDITION FOR OPERATION

3.11.2.4 The gaseous radwaste treatment system and the ventilation exhaust treatment system shall be OPERABLE. The appropriate portions of the gaseous radwaste treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from the site (see Figure 5.1-3), when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The appropriate portions of the ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site (see Figure 5.1-3) when averaged over 31 days would exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the gaseous radwaste treatment system and/or the ventilation exhaust treatment system inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by Specifica- tion 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLE

status, and

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

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days, in accordance with the ODCM.

4.11.2.4.2 The gaseous radwaste treatment system and ventilation exhaust system shall be demonstrated OPERABLE by operating the gaseous radwaste treatment system equipment and ventilation exhaust treatment system equipment for at least minutes, at least once per 92 days unless the appropriate system has been Ttvlized to process radioactive gaseous effluents during the previous 92 days.

PWR-STS- I 3/4 11-14

RADIOACTIVE EFFLUENTS

EXPLOSIVE GAS MIXTURE (Systems designed to withstand a hydrogen explosion)

LIMITING CONDITION FOR OPERATION

3.11.2.5 The concentration of hydrogen or oxygen in the waste gas holdup system shall be limited to less than or equal to 4% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of hydrogen or oxygen in the waste gas holdup system exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of hydrogen or oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously or monitoring the waste gases in the waste gas holdup system with the hydrogen oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10.

PWR-STS-I W3/4 11-15

RADIOACTIVE EFFLUENTS

EXPLOSIVE GAS MIXTURE (Systems not designed to withstand a hydrogen explosion)

LIMITING CONDITION FOR OPERATION

3.11.2.5A The concentration of hydrogen and/or oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of hydrogen and/or oxygen in the waste gas holdup system greater than 2% by volume but less than or equal to 4%

by volume, restore the concentration of hydrogen and/or oxygen to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b. With the concentration of hydrogen and/or oxygen in the waste gas holdup system greater than 4% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of hydrogen and/or oxygen to less than or equal to 2% within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5A The concentrations of hydrogen and/or oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the hydrogen and/or oxygen monitors required OPERABLE by Table 3.3-13 of Specifica- tion 3.3.3.10.

PWR- STS- I 3/4 11-16

RADIOACTIVE EFFLUENTS

GAS STORAGE TANKS

LIMITING CONDITION FOR OPERATION

in each gas storage tank

3.11.2.6 The quantity of radioactivity contained noble gases (considered shall be limited to less than or equal to _ curies as Xe-133).

APPLICABILITY: At all times.

ACTION:

gas storage tank a. With the quantity of radioactive material in any additions of exceeding the above limit, immediately suspend all reduce the tank radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months contents to within the limit.

not applicable.

b. The provisions of Specifications 3.0.3 and 3.0.4 are SURVEILLANCE REQUIREMENTS

in each gas storage

4.11.2.6 The quantity of radioactive material contained limit at least once per 24 tank shall be determined to be within the above the tank.

to hours when radioactive materials are being added PWR-STS-I 3/4 11-17

RADIOACTIVE EFFLUENTS

3/4.11.3 SOLID RADIOACTIVE WASTE

LIMITING CONDITION FOR OPERATION

3.11.3 The solid radwaste system shall be OPERABLE and used, as applicable in accordance with a PROCESS CONTROL PROGRAM, for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site.

APPLICABILITY: At all times.

ACTION:

a. With the packaging requirements of 10 CFR Part 20 and/or 10 CFR Part

71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.

b. With the solid radwaste system inoperable for more than 31 days, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specification

6.9.2 a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLE status,

3. A description of the alternative used for SOLIDIFCATION and packaging of radioactive wastes, and

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

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.3.1 The solid radwaste system shall be demonstrated OPERABLE at least once per 92 days by:

a. Operating the solid radwaste system at least once in the previous

92 days in accordance with the PROCESS CONTROL PROGRAM, or b. Verification of the existence of a valid contract for SOLIDIFCATION

to be performed by a contractor in accordance with a PROCESS CONTROL

PROGRAM.

PWR- STS- I 3/4 11-18

N-'

RADIOACTIVE EFFLUENTS

SURVEILLANCE REQUIREMENTS (Continued)

SOLIDIFICA-

4.11.3.2 THE PROCESS CONTROL PROGRAM shall be used to verify the tenth test specimen from at least every TION of at least one representative spent batch of each type of wet radioactive waste (e.g., filter sludges, solutions).

resins, evaporator bottoms, boric acid solutions, and sodium sulfate a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICA-

TION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION

parameters can be determined in accordance with the PROCESS CONTROL

PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFCATION

of the batch may then be resumed using the alternative SOLIDIFICATION

parameters determined by the PROCESS CONTROL PROGRAM.

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least

3 consecutive initial test specimens demonstrate SOLIDIFCATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste.

PWR-STS-I 3/4 11-19

N-

RADIOACTIVE EFFLUENTS

3/4.11.4 TOTAL DOSE

LIMITING CONDITION FOR OPERATION

3.11.4 The dose or dose commitment to any real individual from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem) over 12 consecutive months.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifica- tions 3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a, 3.11.2.2.b, 3.11.2.3.a, or

3.11.2.3.b, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 and limit the subsequent releases such that the dose or dose commitment to any real individual from uranium fuel cycle sources is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months. This Special Report shall include an analysis which demonstrates that radiation exposures to any real individual from uranium fuel cycle sources (including all effluent pathways and direct radiation) are less than the 40 CFR Part 190 Standard. Otherwise, obtain a variance from the Commission to permit releases which exceed the 40 CFR Part 190

Standard.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2,

4.11.2.2, and 4.11.2.3, and in accordance with the ODCM.

PWR-STS-I 3/4 11-20

I P

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING

3/4.12.1 MONITORING PROGRAM

LIMITING CONDITION FOR OPERATION

conducted

3.12.1 The radiological environmental monitoring program shall be as specified in Table 3.12-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being the conducted as specified in Table 3.12-1, prepare and submit to Operating Report, a description Commission, in the Annual Radiological of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 3.12-2 when averaged over

30

any calendar quarter, prepare and submit to the Commission within days from the end of the affected calendar quarter a Report pursuant to 6.9.1.13. When more than one of the radionuclides in Table be

3.12-2 are detected in the sampling medium, this report shall submitted if:

concentration (1)+ concentration (2)+ > 1.0

limit level (1) limit level (2) -

and When radionuclides other than those in Table 3.12-2 are detected this report shall be submitted if are the result of plant effluents, the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 of and 3.11.2.3. This report is not required if the measured level result of plant effluents; however, in radioactivity was not the the such an event, the condition shall be reported and described in Annual Radiological Environmental Operating Report.

or c. With milk or fresh leafy vegetable samples unavailable from one required by Table 3.12-1, in lieu of more of the sample locations any other report required by Specification 6.9.1, prepare and submit a

to the Commission within 30 days, pursuant to Specification 6.9.2, the cause of the unavailability of Special Report which identifies samples and identifies locations for obtaining replacement samples.

The locations from which samples were unavailable may then be deleted which from those required by Table 3.12-1, provided the locations from the replacement samples were obtained are added to the environmental monitoring program as replacement locations.

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

PWR-STS-I 3/4 12-1

RADIOLOGICAL ENVIRONMENTAL MONITORING

SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the table and figure in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12-1 and 4.12-1.

PWR- STS- I 3/4.12-2

TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

-I

I-

Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations** Collection Frequency of Analysis

1. AIRBORNE

Radioiodine and (Locations 1-5) Continuous operation of Radioiodine canister.

Particulates sampler with sample col- lection as required by Analyze at least once per 7 days for I-131.

(

dust loading but at least once per 7 days. Particulate sampler.

Analyze for gross beta radioactivity > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following filter change.

W Perform gamma isotopic analysis on each sample t-S when gross beta activity I

Wa is > 10 times the yearly mean of control samples.

Perform gamma isotopic analysis on composite (by location) sample at least once per

92 days.

(

2. DIRECT RADIATION (Locations 1-8) At least once per 31 days. Gamma dose. At least

> 2 dosimeters or > 1 once per 31 days.

or or instrument for con- tinuously measuring At least once per 92 days. Gamma dose. At least and recording dose (Read-out frequencies are once per 92 days.

rate at each determined by type of dosi- location. meters selected.)

- Sample locations are given on the figure and table in the ODCM.

TABLE 3.12-1 (Continued)

-o RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

J,I

--I

VW

-I Number of Samples and Sampling and Type and Frequency Exposure Pathway of Analysis and/or Sample Sample Locations" Collection Frequency

3. WATERBORNE

(Locations 9 and 10) Composite* sample collected Gamma isotopic analysis a. Surface of each composite sample.

over a period of < 31 days.

Tritium analysis of com- posite sample at least once per 92 days.

(Locations 11 and 12) At least once per 92 days. Gamma isotopic and b. Ground tritium analyses of each sample.

(Locations 13-15) Composite* sample collected I-131 analysis of each c. Drinking composite sample;

over a period of < 14 days, if I-131 analysis is and performed; or

4- Composite* sample collected Gross beta and gamma over a period of < 31 days. isotopic analysis of each composite sample.

Tritium analysis of composite sample at least once per 92 days.

(Locations 18) At least once per 184 days. Gamma isotopic analysis d. Sediment from of each sample.

Shoreline exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Composite samples shall be collected by collecting an aliquot at intervals not

- Sample locations are shown on the figure in the ODCM.

F

TABLE 3.12-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

-I

IU,

Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations** Collection Frequency of Analysis

4. INGESTION

a. Milk (Locations 17-20) At least once per 15 days Gamma isotopic and ( .i when animals are on pasture; I-131 analysis at least once per 31 days of each sample.

at other times.

b. Fish and (Locations 21 and 22) One sample in season, or at Gamma isotopic analysis Invertebrates least once per 184 days if on edible portions.

not seasonal. One sample of each of the following species:

WA 2.

2.

c. Food Products (Locations 23-25) At time of harvest. One Gamma isotopic analysis sample of each of the fol- on edible portion.

lowing classes of food products:

(

2.

3.

(Location 26) At time of harvest. One I-131 analysis.

sample of broad leaf vegetation.

- Sample locations are shown on the figure in the ODCM.

TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES

-V

Reporting Levels Water Airborne Particulate Fish Milk Food Products Analysis (pCi/l) or Gases (pCi/m 3 ) (pCi/Kg, wet) (pCi/l) (pCi/Kg, wet)

4(a)

H-3 2 x 10 4

a Mn-54 1 x 103 3 x 10

Fe-59 4 x 102 1 x 104 Co-58 1 x 103 3 x 104

-S

r1 Co-60 3 x 102 1 x 104 Zn-65 3 x 102 2 x 104

2(b)

Zr-Nb-95 4 x 10

2 0.9 3 1 x 102 I

I-131 Cs-134 30 10 1 x 103 60 1 x 103 Cs-137 50 20 2 x 103 70 2 x 103 Ba-La-140 2 x 102 3 x1 (a) For drinking water samples. This is 40 CFR Part 141 value.

(b) Total for parent and daughter.

TABLE 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a'd I I I I

-- 4 Airborne Particulate Water or Gag Fish Milk Food Products Sediment Analysis (pCI/1) (pCi/m ) (pCi/kg,wet) (pCi/1) I (pCi/kgwet) (pCi/kg, dry)

II I I I I

gross beta 4b 1 X 102

3H 2000 ( 1 0 0 0 b)

54Mn 15 130

(

5 9 Fe 30 260

5 8 , 6 0 Co

15 130

6 5 Zn w 30 260

'1-a

15 c

9 5 Zr-Nb N

131I lb 7 X 10 2 1 60

134,137Cs 15 (job), 18 1 X 10o 2 130 15 60 150

15 c

1 40 Ba-La 15 c A

I ! . ,

(

TABLE 4.12-1 (Continued)

TABLE NOTATION

a - The LLD is the smallest concentration of radioactive material in a sample that will'be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real"

signal.

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

LLD = 4.66 s E " V " 2.22 " Y " exp(-Aat)

Where:

LLD is the lower limit of detection as defined above (as picocurie per unit mass or volume),

s is the standard deviation of the background counting rate or oY the counting rate of a blank sample as appropriate (as counts per minute),

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

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

2.22 is the number of transformation per minute per picocurie, Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples).

The value of s used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g., potassium-40 in milk samples). Typical values of E, V, Y and At shall be used in the calculations.

PWR-STS-I 3/4 12-8

TABLE 4.12-1 (Continued)

1ABLE NDTATION

b - LLD for drinking water.

c - Total for parent and daughter.

together with the d - Other peaks which are measurable and identifiable, and reported.

radionuclides in Table 4.12-1, shall be identified PWR-STS-I 3/4 12-9

RADIOLOGICAL ENVIRONMENTAL MONITORING

3/4.12.2 LAND USE CENSUS

LIMITING CONDITION FOR OPERATION

3.12.2 A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden* of greater than 500 square feet producing fresh leafy vegetables in each of the

16 meteorological sectors within a distance of five miles. (For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify the locations of all milk animals and all gardens of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of three miles.)

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location(s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3.1, in lieu of any other report required by Specification 6.9.1., prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new location(s).

b. With a land use census identifying a location(s) which yields a calculated dose or dose commitment (via the same exposure pathway)

greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new location. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location, excluding the control station location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted at least once per 12 months between the dates of (June 1 and October 1) by a door-to-door survey, aerial survey, or by consulting local agriculture authorities, using that information which will provide the best results.

Broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census.

PWR-STS-I 3/4 12-10

RADIOLOGICAL ENVIRONMENTAL MONITORING

3/4.12.3 INTERLABORATORY COMPARISON PROGRAM

LIMITING CONDITION FOR OPERATION

3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission.

APPLICABILITY: At all times.

ACTION:

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

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the ODCM shall be included in the Annual Radiological Environmental Operating Report.

PWR- STS- I 3/4 12-11

INSTRUMENTATION

BASES

3/4.3.3.9 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION

to monitor The radioactive liquid effluent instrumentation is provided in liquid materials and control, as applicable, the releases of radioactive effluents. The alarm/

effluents during actual or potential releases of liquid in accordance with trip setpoints for these instruments shall be calculated will occur prior to the procedures in the ODCM to ensure that the alarm/trip and use of this exceeding the limits of 10 CFR Part 20. The OPERABILITY General Design Criteria instrumentation is consistent with the requirements of

60, 63 and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.10 RADIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION

to monitor The radioactive gaseous effluent instrumentation is provided in gaseous materials and control, as applicable, the releases of radioactive effluents. The effluents during actual or potential releases of gaseous in accordance alarm/trip setpoints for these instruments shall be calculatedwill occur prior with the procedures in the ODCM to ensure that the alarm/trip also includes to exceeding the limits of 10 CFR Part 20. This instrumentationof potentially provisions for monitoring (and controlling) the concentrationsThe OPERABILITY and explosive gas mixtures in the waste gas holdup system. of General use of this instrumentation is consistent with the requirements 50.

Part Design Criteria 60, 63 and 64 of Appendix A to 10 CFR

PWR-STS- I B 3/4 3-5

- J

3/4.11 RADIOACTIVE EFFLUENTS

BASES

3/4.11.1 LIQUID EFFLUENTS

3/4.11.1.1 CONCENTRATION

of radio- This specification is provided to ensure that the concentration will be less waste effluents from the site active materials released in liquid Appendix B, Table than the concentration levels specified in 10 CFR Part 20,

additional assurance that the levels II, Column 2. This limitation provides result in of radioactive materials in bodies of water outside the site will I, 10 CFR

exposures within (1) the Section II.A design objectives of Appendix of 10 CFR 20.106(e) to the population.

50, to an individual and (2) the limits is based upon The concentration limit for dissolved or entrained noble gases its MPC in air the assumption that Xe-135 is the controlling radioisotope and using the (submersion) was converted to an equivalent concentration in water Protection methods described in International Commission on Radiological (ICRP) Publication 2.

3/4.11.1.2 DOSE

of Sections This specification is provided to implement the requirements Condition II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting of Appendix I.

for Operation implements the guides set forth in Section II.A and at the The ACTION statements provide the required operating flexibility IV.A of Appendix I to same time implement the guides set forth in Section will be assure that the releases of radioactive material in liquid effleunts fresh water sites with kept "as low as is reasonably achievable". Also, for plant operations, drinking water supplies which can be potentially affected by will not there is reasonable assurance that the operation of the facility drinking water that are result in radionuclide concentrations in the finished in the in excess of the requirements of 40 CFR 141. The dose calculations of Appendix I that conformance ODCM implement the requirements in Section III.A based on with the guides of Appendix I be shown by calculational procedures through appro- models and data, such that the actual exposure of an individual underestimated. The equations priate pathways is unlikely to be substantially release specified in the ODCM for calculating the doses due to the actual with the rates of radioactive materials in liquid effluents are consistent of Annual Doses methodology provided in Regulatory Guide 1.109, "Calculation of Evaluating to Man from Routine Releases of Reactor Effluents for the

Purpose

Revision 1, October 1977 and Compliance with 10 CFR Part 50, Appendix I," from Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents Accidental and Routine Reactor Releases for the

Purpose

of Implementing Appendix I," April 1977.

each This specification applies to the release of liquid effluents from the shared radwaste treatment systems, reactor at the site. For units with the units liquid effluents from the shared system are proportioned among sharing that system.

PWR-STS-I B 3/4 11-1

RADIOACTIVE EFFLUENTS

BASES

3/4.11.1.3 LIQUID WASTE TREATMENT

The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents 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 of radioactive materials in liquid effluents will be 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 the design objective given in 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 a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

3/4.11.1.4 LIQUID HOLDUP TANKS

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of

10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.

3/4.11.2 GASEOUS EFFLUENTS

3/4.11.2.1 DOSE RATE

This specification is provided to ensure that the dose at any time at the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20,

Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)(1)).

For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem/ year to the total body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to less than or equal to 1500 mrem/

year for the nearest cow to the plant.

PWR-STS-I B 3/4 11-2

RADIOACTIVE EFFLUENTS

BASES

from all This specification applies to the release of gaseous effluents the treatment systems, reactors at the site. For units with shared radwaste units among the gaseous effluents from the shared system are proportioned sharing that system.

3/4.11.2.2 DOSE, NOBLE GASES

of Sections II.B,

This specification is provided to implement the requirements for Limiting Condition III.A and IV.A of Appendix I, 10 CFR Part 50. The I. The II.B of Appendix Operation implements the guides set forth in Section at the same flexibility and ACTION statements provide the required operating I to assure IV.A of Appendix time implement the guides set forth in Section will be kept that the releases of radioactive material in gaseous effluents implement Requirements

"as low as is reasonably achievable". The Surveillance with the the requirements in Section III.A of Appendix I that conformance procedures based on models and guides of Appendix I be shown by calculational pathways through appropriate data such that the actual exposure of an individual calculations established is unlikely to be substantially underestimated. The dose actual release rates of in the ODCM for calculating the doses due to the the methodology are consistent with radioactive noble gases in gaseous effluents to Man from of Annual Doses provided in Regulatory Guide 1.109, "Calculation Compliance

Purpose

of Evaluating Routine Releases of Reactor Effluents for the Regulatory

1, October 1977 and with 10 CFR Part 50, Appendix I," Revision and Dispersion of Guide 1.111, "Methods for Estimating Atmospheric Transport Light-Water Cooled Reactors,"

Gaseous Effluents in Routine Releases from the air provided for determining Revision 1, July 1977. The ODCM equations atmospheric the historical average doses at the site boundary are based upon conditions.

IN PARTICULATE FORM

3/4.11.2.3 DOSE, RADIOIODINES, RADIOACTIVE MATERIALS

AND RADIONUCLIDES OTHER THAN NOBLE GASES

of Sections II.C,

This specification is provided to implement the requirements Condition for Opera- III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting II.C of Appendix I. The ACTION state- tion are the guides set forth in Section time implement flexibility and at the same ments provide the required operating the releases of Appendix I to assure that the guides set forth in Section IV.A as is effluents will be kept "as low of radioactive materials in gaseous in the calculational methods specified reasonably achievable". The ODCM III.A of the requirements in Section Surveillance Requirements implement by calcula- I be shown Appendix I that conformance with the guides of Appendix actual exposure of an tional procedures based on models and data, such that the substantially individual through appropriate pathways is unlikely to be methods for calculating the doses due underestimated. The ODCM calculational consistent with the to the actual release rates of the subject materials are of Annual Doses methodology provided in Regulatory Guide 1.109, "Calculation PWR-STS-I B 3/4 11-3

-

RADIOACTIVE EFFLUENTS

BASES

to Man from 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 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions.

The release rate specifications for radioiodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consump- tion of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

3/4.11.2.4 GASEOUS RADWASTE TREATMENT

The OPERABILITY of the gaseous radwaste treatment system and the ventila- tion exhaust treatment system ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be 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 the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

3/4.11.2.5 EXPLOSIVE GAS MIXTURE

This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. (Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits.) Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

PWR-STS-I B 3/4 11-4

Ig RADIOACTIVE EFFLUENTS

BASES

3/4.11.2.6 GAS STORAGE TANKS

Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, "Waste Gas System Failure".

3/4.11.3 SOLID RADIOACTIVE WASTE

The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL

PROGRAM may include, but are not limited to waste type, waste pH, waste/liquid/

solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

3/4.11.4 TOTAL DOSE

This specification is provided to meet the dose limitations of 40 CFR 190.

The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a real individual will exceed

40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to a real Individual for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the real individual 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 5 miles must be considered.

PWR-STS-I B 3/4 11-5

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING

BASES

3/4.12.1 MONITORING PROGRAM

by this specification provides The radiological monitoring program required materials in those exposure measurements of radiation and of radioactive lead to the highest potential pathways and for those radionuclides, which from the station operation. This radiation exposures of individuals resulting the radiological effluent monitoring monitoring program thereby supplements concentrations of radioactive materials program by verifying that the measurable expected on the basis of the and levels of radiation are not higher than environmental exposure pathways.

effluent measurements and modeling of the will be effective for at least the The initially specified monitoring program Following this period, program first three years of commercial operation. experience.

changes may be initiated based on operational

4.12-1 are state-of-the-art The detection capabilities required by Table laboratories. The LLD's in industrial for routine environmental measurements of 40 CFR 141.

for drinking water meet the requirements

3/4.12.2 LAND USE CENSUS

that changes in the use of This specification is provided to ensure modifications to the monitoring unrestricted areas are identified and that of this census. The best survey program are made if required by the results consulting with local agricultural or information from the door-to-door, aerial the requirements of Section satisfies authorities shall be used. This census Restricting the census to gardens of IV.B.3 of Appendix I to 10 CFR Part 50. that significant exposure greater than 500 square feet provides assurance and monitored since a garden pathways via leafy vegetables will be identified the quantity (26 kg/year) of produce of this size is the minimum required to 1.109 for consumption by a child.

Guide leafy vegetables assumed in Regulatory following assumptions were used, 1)

To determine this minimum garden size, the broad leaf vegetation (i.e.,

that 20% of the garden was used for growing a vegetation yield of 2 kg/square similar to lettuce and cabbage), and 2)

meter.

3/4.12.3 INTERLABORATORY COMPARISON PROGRAM

Interlaboratory Comparison Program The requirement for participation in an on the precision and accuracy of is provided to ensure that independent checks in environmental sample matrices are the measurements of-radioactive material program for environmental monitoring performed as part of the quality assurance are reasonably valid.

in order to demonstrate that the results PWR-STS-I B 3/4 12-1

5.0 DESIGN FEATURES

5.1 SITE

EXCLUSION AREA

5.1.1 The exclusion area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE

5.1.2 The low population zone shall be as shown in Figure 5.1-2.

SITE BOUNDARY FOR GASEOUS EFFLUENTS

5.1.3 The site boundary for gaseous effluents shall be as shown in Figure 5.1-3.

SITE BOUNDARY FOR LIQUID EFFLUENTS

5.1.4 The site boundary for liquid effluents shall be as shown in Figure 5.1-4.

5.2 CONTAINMENT

CONFIGURATION

5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape,-with a dome roof and having the following design features:

a. Nominal inside diameter = feet.

b. Nominal inside height = feet.

c. Minimum thickness of concrete walls = feet.

d. Minimum thickness of concrete roof = feet.

e. Minimum thickness of concrete floor pad = feet.

f. Nominal thickness of steel liner = inches.

g. Net free volume = cubic feet.

DESIGN PRESSURE AND TEMPERATURE

5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of psig and a temperature of OF.

PWR- STS- I .5-1

This figure shall consist of a map of the site area and provide, at a minimum, the information described in Section (2.1.2) of the FSAR and meteorological tower location.

EXCLUSION AREA

FIGURE 5.1-1 PWR-STS-I 5-2

This figure shall consist of a map of the site area showing the Low Population Zone boundary.

Features such as towns, roads and recreational areas shall be indicated in sufficient detail to allow identification of significant shifts in population distribution within the LPZ.

LOW POPULATION ZONE

FIGURE 5.1-2 PWR-STS- I 5-3

- -

This figure shall consist of a map of the site area showing the perimeter of the site and locating the points where gaseous effluents are released. If on-site land areas subject to radioactive materials in gaseous waste are utilized by the public for recreational or other purposes, then these areas shall be identified by occupancy factors and the licensee's method of occupancy control. The figure shall be sufficiently detailed to allow identification of structures and release point elevations, and areas within the site boundary that are accessible by members of the general public. See NUREG-0133 for additional guidance.

SITE BOUDARY FOR GASEOUS EFFLUENTS

FIGURE 5.1-3 PWR-STS-I 5-4

a This figure shall consist of a map of the site area showing the perimeter of the site and locating the points where liquid effluent leaves the site. If on-site water areas containing radioactive wastes are utilized by the public for recreational or other purposes, the points of release to these water areas shall be identified. The figure shall be sufficiently detailed to allow identifica- tion of structures near the release points and areas within the site boundary where ground and surface water is accessible by members of the general public. See NUREG-0133 for additional guidance.

SITE BOUNDARY FOR LIQUID EFFLUENTS

FIGURE 5.1-4 PWR-STS-I 5-5

a

2 ALL STS-I

SECTION 6.0

ADMINISTRATIVE CONTROLS

6.0 ADMINISTRATIVE CONTROLS

6.5.1 UNIT REVIEW GROUP (URG)

RESPONSIBILITIES

6.5.1.6 The URG shall be responsible for:

k. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence to the (Superintendent of Power Plants)

and to the (Company Nuclear Review and Audit Group).

1. Review of major changes to radwaste systems.

6.5.2 COMPANY NUCLEAR REVIEW AND AUDIT GROUP (CNPAG)

AUDITS

6.5.2.8 Audits of unit activities shall be performed under the cognizance of the (CNRAG). These audits shall encompass:

1. The radiological environmental monitoring program and the results thereof at least once per 12 months.

m. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

n. The PROCESS CONTROL PROGRAM and implementing procedures for solidifica- tion of radioactive wastes at least once per 24 months.

O. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, December 1977 at least once per 12 months.

6.8 PROCEDURES

6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 4.15.

ALL STS-I 6-1

ADMINISTRATIVE CONTROLS

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORTW/

6.9.1.6 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality.

6.9.1.7 The annual radiological environmental operating reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, 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 land use censuses required by Specification

3.12.2. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

The annual radiological environmental operating reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period. In the event that some 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 also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.12.3.

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT='

6.9.1.8 Routine radioactive effluent release reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the data of initial criticality.

3/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station;

however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

ALL STS-I 6-2

S

ADMINISTRATIVE CONTROLS

6.9.1.9 The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radio- active Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The radioactive effluent release report to be submitted 60 days after January 1 of each year and shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, 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 released from the unit or station during the previous calendar year. This same report shall also include-an assessment of the radiation doses from radioactive liquid and gaseous effluents to individuals due to their activities inside the site boundary (Figures 5.1-3 and 5.1-4) 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 in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (ODCM).

The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with

40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.

The radioactive effluent release reports shall include the following information for each type of solid waste shipped offsite 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),

ALL STS-I 6-3

ADMINISTRATIVE CONTROLS

d. type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),

e. type of container (e.g., LSA, Type A, Type B, Large Quantity), and f. solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period.

MONTHLY REACTOR OPERATING REPORT

6.9.1.10 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.

Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change(s) was made effective. In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by the (Unit Review Group).

PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP

6.9.1.12 j. Offsite releases of radioactive materials in liquid and gaseous effluents which exceed the limits of Specification 3.11.1.1 or

3.11.2.1.

k. Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for the storage of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits.

ALL STS-I 6-4

0 #F

ADMINISTRATIVE CONTROLS

THIRTY DAY WRITTEN REPORTS

6.9.1.13 e. An unplanned offsite release of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05. curies of radioiodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information:

1. A description of the event and equipment involved.

2. Cause(s) for the unplanned release.

3. Actions taken to prevent recurrence.

4. Consequences of the unplanned release.

f. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 3.12-2 when averaged over any calendar quarter sampling period.

6.10 RECORD RETENTION

6.10.2

1. Records of analyses required by the radiological environmental monitoring program.

ALL STS-I 6-5

ADMINISTRATIVE CONTROLS

6.13 PROCESS CONTROL PROGRAM (PCP)

6.13.1 The PCP shall be approved by the Commission prior to implementation.

6.13.2 Licensee initiated changes to the PCP:

1. Shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:

a. sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

b. a determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and c. documentation of the fact that the change has been reviewed and found acceptable by the (URG).

2. Shall become effective upon review and acceptance by the (URG).

6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)

6.14.1 The ODCM shall be approved by the Commission prior to implementation.

6.14.2 Licensee initiated changes to the ODCM:

1. Shall be submitted to the Commission in the Monthly Operating Report within 90 days of the date the change(s) was made effective. This submittal shall contain:

a. sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change(s);

b. a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations;

and c. documentation of the fact that the change has been reviewed and found acceptable by the (URG).

2. Shall become effective upon review and acceptance by the (URG).

ALL STS-I 6-6

.,--I-- - --- 6 ADMINISTRATIVE CONTROLS

6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and solid)

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

1. Shall be reported to the Commission in the Monthly Operating Report for the period in which the evaluation was reviewed by the (Unit Review Group). 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 50.59;

b. sufficient 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. an evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

e. an evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto;

f. 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;

g. an estimate of the exposure to plant operating personnel as a result of the change; and h. documentation of the fact that the change was reviewed and found acceptable by the (URG).

2. Shall become effective upon review and acceptance by the (URG).

ALL STS-I 6-7

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