NRC Generic Letter 1979-30

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

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 3/4.11 RADIOACTIVE

EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................

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

Liquid Waste Treatment...................................

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

PAGE 3/4 3/4 3/4 3/4 11-1 11-5 11-6 11-7 3/4 11.2 GASEOUS EFFLUENTS Dose Rate................................................

Dose-Noble Gases.........................................

Dose-Radioiodines, Particulate, and Radionuclides Other than Noble Gases...................

Gaseous Waste Treatment..................................

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

Gas Storage Tanks........................................

3/4 11.3 SOLID RADIOACTIVE

WASTE..................................

3/4 11.4 TOTAL DOSE...............................................

3/4.12 RADIOLOGICAL

ENVIRONMENTAL

MONITORING

3/4.12.1 MONITORING

PROGRAM......................................

3/4.12.2 LAND USE CENSUS.........................................

3/4.12.3 INTERLABORATORY

COMPARISON..............................

3/4 11-8 3/4 11-12 3/4 3/4 3/4 3/4 11-13 11-14 11-15 11-17 3/4 11-18 3/4 11-20 3/4 3/4 3/4 12-1 12-10 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

1.9 A CHANNEL CALIBRATION

shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRA-TION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL

TEST. The CHANNEL CALIBRATION

may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indica-tions and/or status derived from independent instrumentation channels measuring 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 isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present.The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites." SOURCE CHECK 1.29 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.PROCESS CONTROL PROGRAM (PCP)1.30 The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION

of radioactive wastes from 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 other operating condition, in such a manner that replacement air or gas is not 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 a. Liquid Radwaste Effluent Line (1) 28 b. Steam Generator Blowdown Effluent Line (1) 29 c. Turbine Building (Floor Drains) Sumps Effluent Line (1) 30 2. GROSS RADIOACTIVITY

MONITORS NOT PROVIDING

AUTOMATIC v ~ TERMINATION

OF RELEASE a. Service Water System Effluent Line (1) 30 cm b. Component Cooling Water System Effluent Line (1) 30 3. CONTINUOUS

COMPOSITE

SAMPLERS AND SAMPLER FLOW MONITOR a. Steam Generator Blowdown Effluent Line (1) 29 b. Turbine Building Sumps Effluent Line (1) 30 4. FLOW RATE MEASUREMENT

DEVICES a. Liquid Radwaste Effluent Line (1) 31 b. Discharge Canal (1) 31 c. Steam Generator Blowdown Effluent Lines (1) 31

.TABLE 3.3-12 (Continued)

RADIOACTIVE

LIQUID EFFLUENT MONITORING

INSTRUMEN

rATION I-MINIMUM CHANNELS OPERABLE INSTRUMENT

5. RADIOACTIVITY

RECORDERS

(*)a. Liquid Radwaste Effluent Line b. Steam Generator Blowdown Effluent Line 6. TANK LEVEL INDICATING

DEVICES (for tanks outside plant buildings)

a.b.C.ACTION (1).(1)33 34'I-(1)(1)(1)(1)32 32 32 32 d.((xRequired only if alarm/trip set point is based on recorder-controller)

TABLE 3.3-12 (Continued)

TABLE NOTATION ACTION 28 -ACTION 29 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may 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.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 analyzed for gross radioactivity (beta or gamma) at 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.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 that, at 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.

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.With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue for up to 30 days provided the tank liquid level is estimated during all liquid additions to the tank.ACTION 30 -ACTION 31 -ACTION 32 -3/4 3-68 TABLE 3.3-12 (Continued)

TABLE NOTATION ACTION 33 -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 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.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 Ln RADIOACTIVE

LIQUID EFFLUENT MONITORING

INSTRUMENTATION

SURVEILLANCE

REQUIREMENTS

CHANNEL CHECK SOURCE CHECK CHANNEL CALIBRATION

CHANNEL FUNCTIONAL

TEST INSTRUMENT

1. GROSS BETA OR GAMMA RADIOACTIVITY

MONITORS PROVIDING

ALARM AND AUTOMATIC

TERMINATION

OF RELEASE a. Liquid Radwaste Effluents Line b. Steam Generator Blowdown Effluent Line c. Turbine Building (Floor Drains) Sumps Effluent Line (D D D P M M R(3)R(3)R(3)Q(M)Q(M)Q(M)w w 0 2.GROSS BETA OR GAMMMA RADIOACTIVITY

MONITORS PROVIDING

ALARM BUT NOT PROVIDING

AUTOMATIC TERMINATION

OF RELEASE a. Service Water System Effluent Line b. Component Cooling Water System Effluent Line D M R(3)R(3)Q(2)D M Q(2)(3. CONTINUOUS

COMPOSITE

SAMPLERS AND SAMPLER FLOW MONITOR a. Steam Generator Blowdown Effluent Line b. Turbine Building Sumps Effluent Line D N. A.R R Q Q 0 N.A.it TABLE 4.3-12 (Continued)

RADIOACTIVE

LIQUID EFFLUENT MONITORING

INSTRUMENTATION

SURVEILLANCE

REQUIREMENTS

-0 Z-M--4 LA I INSTRUMENT

4. FLOW RATE MEASUREMENT

DEVICES a. Liquid Radwaste Effluent Line b. Steam Generator Blowdown Effluent Line c. Discharge Canal CHANNEL CHECK D(4)D(4)D(4)SOURCE CHECK N.A.N.A.N. A.CHANNEL CALIBRATION

R R R CHANNEL FUNCTIONAL

TEST Q Q Q C 5. RADIOACTIVITY

RECORDERS 4 a. Steam Generator Blowdown Effluent Line w b. Liquid Radwaste Effluent Line 6. TANK LEVEL INDICATING

DEVICES (for tanks outside the building)a.b.C.d.0 0 N. A.N. A.R R Q Q D*D*D*D*N. A.N. A.N. A.N. A.R R R R Q Q Q Q

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

4j_TABLE 3.3-13 RADIOACTIVE

GASEOUS EFFLUENT MONITORING

INSTRUMENTATION

MINIMUM CHANNELS nrnADI C INSTRUMENT

ur 1. WASTE GAS HOLDUP SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Effluent System Flow Rate Measuring Device e. Sampler Flow Rate Measuring Device 2A. WASTE GAS HOLDUP SYSTEM EXPLOSIVE

GAS MONITORING

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

a. Hydrogen Monitor b. Hydrogen or Oxygen Monitor'EMMULL APPLICABILITY

ACTION (1)(1)*(1)(1)(1)**-0 4A)~4 35 41 41 36 36 39 39 (1)(1)**2B. WASTE GAS HOLDUP SYSTEM EXPLOSIVE

GAS MONITORING

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

a. Hydrogen Monitor b. Hydrogen or Oxygen Monitor (2)(2)**40 40

TABLE 3.3-13 (Continued)

RADIOACTIVE

GASEOUS EFFLUENT MONITORING

INSTRUMENTATION

INSTRUMENT

3. CONDENSER

EVACUATION

SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor MINIMUM CHANNELS OPERABLE (1)-(1)(1)(1)(1)APPLICABI

LITY*I**ACTION 37 41 41 36 36 (--j C", 4. VENT a.b.c.d.e.HEADER SYSTEM Noble Gas Activity Monitor Iodine Sampler Particulate Sampler Flow Rate Monitor Sampler Flow Rate Monitor (1)(1)(1)(1)(1)*****37 41 41 36 36 5. CONTAINMENT

PURGE SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler (1)(1)(1)**38 41 41

-o:C (A TABLE 3.3-13 (Continued)

RADIOACTIVE

GASEOUS EFFLUENT MONITORING

INSTRUMENTATION

INSTRUMENT

5. CONTAINMENT

PURGE SYSTEM (Continued)

e. Flow Rate Monitor f. Sampler Flow Rate Monitor MINIMUM CHANNELS OPERABLE (1)(1)APPLICABILITY

- ACTION 36 36 (4aP.c4M 6. AUXILIARY

BUILDING VENTILATION

SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor (1)(1)(1)(1)(1)*****37 41 41 36 36 7. FUEL a.b.c.d.e.STORAGE AREA VENTILATION

SYSTEM Noble Gas Activity Monitor Iodine Sampler Particulate Sampler Flow Rate Monitor Sampler Flow Rate Monitor (1)(1)(1)(1)(1)*****37 41 41 36 36

-o-94 (U, TABLE 3.3-13 (Continued)

RADIOACTIVE

GASEOUS EFFLUENT MONITORING

INSTRUMENTATION

l INSTRUMENT

8. RADWASTE AREA VENTILATION

SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor 9. STEAM GENERATOR

BLOWDOWN VENT SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor MINIMUM CHANNELS OPERABLE (1)(1)(1)(1)(1)(1)(1)(1)(1)(1)APPLICABILITY

- - - A 37 41 41 36 36 (ACTION*37 41 41 36 36

4 TABLE 3.3-13 (Continued)

TABLE NOTATION* At all times.** During waste gas holdup system operation (treatment for primary system offgases).

ACTION 35 -ACTION 36 -ACTION 37 -ACTION 38 -ACTION 39 -ACTION 40 -ACTION 41 -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.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 .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 .With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.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 .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 .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 MONITORING

INSTRUMENTATION

SURVEILLANCE

REQUIREMENTS

a--i (A I-RAnTDArTTVF

GASEOUS EFFLUENT INSTRUMENT

1. WASTE GAS HOLDUP SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor CHANNEL CHECK P WI W P D SOURCE CHECK P N. A.N. A.N. A.N. A.CHANNEL CALIBRATION

R(3)N.A.N.A.R R CHANNEL FUNCTIONAL

TEST Q(l)N.A.N.A.Q Q MODES IN WHICH SURVEILLANCE

REQUIRED***(w.bi t 2.WASTE GAS HOLDUP SYSTEM EXPLOSIVE GAS MONITORING

SYSTEM a. Hydrogen Monitor b. Hydrogen Monitor (alternate)

c. Oxygen Monitor d. Oxygen Monitor (alternate)

D 0 D D N/A N/A N/A N.A.Q(4)Q(4)Q(5)Q(5)M M M M******** (j

-I (A~TABLE 4.3-13 (Continued)

RADIOACTIVE

GASEOUS EFFLUENT MONITORING

INSTRUMENTATION

SURVEILLANCE

REQUIREMENTS

la 4b l co CD INSTRUMENT

3. CONDENSER

EVACUATION

SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor 4. VENT HEADER SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor CHANNEL CHECK D W W D D D W W D D SOURCE CHECK M N. A.N. A.N. A.N. A.M N.A.N. A.N. A.N. A.CHANNEL CALIBRATION

R(3)N.A.N.A.R R R(3)N.A.N.A.R R CHANNEL FUNCTIONAL

TEST Q(2)N.A.N.A.Q Q Q(2)N.A.N.A.Q Q MODES IN WHICH SURVEILLANCE

REQUIRED x*******((

-o-I TABLE 4.3-13 (Continued)

RADIOACTIVE

GASEOUS EFFLUENT MONITORING

INSTRUMENTATION

SURVEILLANCE

REQUIREMENTS

1-~'A, CA, IA CHANNEL INSTRUMENT

CHECK 5. CONTAINMENT

PURGE SYSTEM a. Noble Gas Activity Monitor D b. Iodine Sampler W c. Particulate Sampler W d. Flow Rate Monitor D e. Sampler Flow Rate Monitor D 6. AUXILIARY

BUILDING VENTILATION

SYSTEM a. Noble Gas Actvity Monitor D b. Iodine Sampler W c. Particulate Sampler W d. Flow Rate Monitor D e. Sampler Flow Rate Monitor D SOURCE CHECK P N. A.N.A.N. A.N. A.M N. A.N. A.N. A.N. A.CHANNEL CALIBRATION

R(3)N.A.N.A.R R R(3)N.A.N.A.R R CHANNEL FUNCTIONAL

TEST Q(l)N.A.N.A.Q Q Q(2)N.A.N.A.Q Q MODES IN WHICH SURVEILLANCE

REQUIRED*******(7. FUEL STORAGE AREA VENTILATION

SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler D W W M N. A.N. A.R(3)N. A.N. A.Q(2)N. A.N. A.A A*

-vi-I TABLE 4.3-13 (Continued)

RADIOACTIVE

GASEOUS EFFLUENT MONITORING

INSTRUMENTATION

SURVEILLANCE

REQUIREMENTS

CHANNEL SOURCE INSTRUMENT

CHECK CHECK 7. FUEL STORAGE AREA VENTILATION

SYSTEM (Continued)

d. Flow Rate Monitor D N.A.e. Sampler Flow Rate Monitor D N.A.CHANNEL CALIBRATION

CHANNEL FUNCTIONAL

TEST Q I Q t MODES IN WHICH SURVEILLANCE

REQUIRED*R R (4-04I 8. RADWASTE AREA VENTILATION

SYSTEM a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor 9. STEAM GENERATOR

BLOWDOWN VENT a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor D W W D D M N. A.N. A.N. A.N. A.R(3)N.A.N. A.R R Q(2)N. A N. A Q Q A**A*(D W W D D M N. A.N. A.N. A.N. A.R(3)N. A.N. A.R R Q(2)N. A N. A.Q Q A*A*A

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)P M H-3 1x10 5 Each Batch Compositeb Gross Alpha x10 7 7 P-32 lx10-6 P Q Sr-89, Sr-90 5x10 8 Each Batch Composite

6 Fe-55 x10-6 B. Continuous W Principal Gamma 5xlO 7 Releasese Continuousc Composite Emitters I-131 1x10-6 M M Dissolved and lxlO 5 Grab Sample Entrained Gases (Gamma Emitters)M H-3 lxlO 5 Continuous c Compositec

7 Gross Alpha 1x1O0 P-32 lxl0 6 Q Sr-89, Sr-90 5xlO 8 ContinuousC

c 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 tanks shall be limited dissolved or entrained of radioactive material contained in each of the following to less than or equal to curies, excluding tritium and 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 (A~-.4 c.F*I to4 RADIOACTIVE

GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Sampling Analysis Type of Detection (hLD)Gaseous Release Type Frequency Frequency Activity Analysis (zCi/m1)_ P PP A. Waste Gas Storage Each Tank Each Tank Principal Gamma Emittersg x10 4 Tank Grab Sample B.Cnaimn Preb b 9-B. Containment Purge Each Purge Each Purge Principal Gamma Emitters x10 4 Grab -Sample H-3 lxlO C. (List other release Mb,ce Mb Principal Gamma Emittersg lxl0&4 points where gas- Grab eous effluents are Sample H-3 lx1 6 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 Continuousf M Gross Alpha lxlO 1 1 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 or thermal power level change exceeding

15% of RATED THERNAL POWER in one hour. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the cor-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 greater than 8 days in gaseous effluents released from the site (see Figure 5.1-3)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 monitoring the waste gases in the waste gas holdup system with the hydrogen or 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 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to _ curies noble gases (considered as Xe-133).APPLICABILITY:

At all times.ACTION: a. With the quantity of radioactive material in any gas storage tank 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

4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.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)

4.11.3.2 THE PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICA-

TION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

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

At all times.ACTION: a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Operating Report, a description 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 any calendar quarter, prepare and submit to the Commission within 30 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 3.12-2 are detected in the sampling medium, this report shall be submitted if: concentration

(1) + concentration

(2) + > 1.0 limit level (1) limit level (2) -When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if 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 and 3.11.2.3.

This report is not required if the measured level of radioactivity was not the result of plant effluents;

however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 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 cause of the unavailability of samples and identifies locations for obtaining replacement samples.The locations from which samples were unavailable may then be deleted from those required by Table 3.12-1, provided the locations from which 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

-I I-TABLE 3.12-1 RADIOLOGICAL

ENVIRONMENTAL

MONITORING

PROGRAM Exposure Pathway and/or Sample Number of Samples and Sample Locations**

Sampling and Collection Frequency Type and Frequency of Analysis 1. AIRBORNE Radioiodine and Particulates (Locations

1-5)Continuous operation of sampler with sample col-lection as required by dust loading but at least once per 7 days.W t-S I Wa Radioiodine canister.Analyze at least once per 7 days for I-131.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.Perform gamma isotopic analysis on each sample when gross beta activity 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)> 2 dosimeters or > 1 instrument for con-tinuously measuring and recording dose rate at each location.At least once per 31 days.or At least once per 92 days.(Read-out frequencies are determined by type of dosi-meters selected.)

Gamma dose. At least once per 31 days.or Gamma dose. At least once per 92 days.** Sample locations are given on the figure and table in the ODCM.

TABLE 3.12-1 (Continued)-o J,I--I VW-I RADIOLOGICAL

ENVIRONMENTAL

MONITORING

PROGRAM Exposure Pathway and/or Sample Number of Samples and Sample Locations" Sampling and Collection Frequency Type and Frequency of Analysis 3. WATERBORNE

a. Surface b. Ground (Locations

9 and 10)(Locations

11 and 12)Composite*

sample collected over a period of < 31 days.At least once per 92 days.Gamma isotopic analysis of each composite sample.Tritium analysis of com-posite sample at least once per 92 days.Gamma isotopic and tritium analyses of each sample.c. Drinking (Locations

13-15)Composite*

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

or 4-Composite*

sample collected over a period of < 31 days.I-131 analysis of each composite sample;and Gross beta and gamma isotopic analysis of each composite sample.Tritium analysis of composite sample at least once per 92 days.d. Sediment from Shoreline (Locations

18)At least once per 184 days.Gamma isotopic analysis of each sample.* Composite samples shall be collected by collecting an aliquot at intervals not exceeding

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .**Sample locations are shown on the figure in the ODCM.

F TABLE 3.12-1 (Continued)-I IU , RADIOLOGICAL

ENVIRONMENTAL

MONITORING

PROGRAM Exposure Pathway and/or Sample Number of Samples and Sample Locations**

Sampling and Collection Frequency Type and Frequency of Analysis 4. INGESTION a. Milk (Locations

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

21 and 22)One sample in season, or at least once per 184 days if not seasonal.

One sample of each of the following species: Gamma isotopic analysis on edible portions.WA 2.2.c.Food Products (Locations

23-25)At time of harvest. One sample of each of the fol-lowing classes of food products: Gamma isotopic analysis on edible portion.(2.3.(Location

26)At time of harvest. One sample of broad leaf vegetation.

I-131 analysis.**Sample locations are shown on the figure in the ODCM.

TABLE 3.12-2 REPORTING

LEVELS FOR RADIOACTIVITY

CONCENTRATIONS

IN ENVIRONMENTAL

SAMPLES-V-S r1;a 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 Mn-54 1 x 103 3 x 10 Fe-59 4 x 102 1 x 104 Co-58 1 x 103 3 x 104 Co-60 3 x 102 1 x 104 Zn-65 3 x 102 2 x 104 2(b)Zr-Nb-95 4 x 10 I-131 2 0.9 3 1 x 102 Cs-134 30 10 1 x 10 3 60 1 x 10 3 Cs-137 50 20 2 x 10 3 70 2 x 10 3 Ba-La-140

2 x 102 3 x 1 (a) For drinking water samples. This is 40 CFR Part 141 value.(b) Total for parent and daughter.I

--4 TABLE 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a'd I I I I Analysis Water (pCI/1)Airborne Particulate or Gag (pCi/m )Fish (pCi/kg,wet)

Milk (pCi/1)Food Products I (pCi/kgwet)

Sediment (pCi/kg, dry)II I I I I gross beta 3 H 54Mn 5 9 Fe 5 8 , 6 0 Co 6 5 Zn 9 5 Zr-Nb 131I 134,137Cs 1 40 Ba-La 4 b 2000 (1 0 0 0 b)15 30 15 30 1 5 c lb 15 (job), 18 1 5 c 1 X 102 (w'1-a N 130 260 130 260 130 7 X 10 2 1 X 10o 2 1 15 1 5 c 60 60 150 I! ., A (

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.d -Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1, shall be identified and reported.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

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.

The alarm/trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY

and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.3/4.3.3.10

RADIOACTIVE

GASEOUS EFFLUENT INSTRUMENTATION

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm/trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling)

the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The OPERABILITY

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

This specification is provided to ensure that the concentration of radio-active materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to an individual and (2) the limits of 10 CFR 20.106(e)

to the population.

The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion)

was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication

2.3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I.The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effleunts will be kept "as low as is reasonably achievable".

Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appro-priate pathways is unlikely to be substantially underestimated.

The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses 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 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the

Purpose

of Implementing Appendix I," April 1977.This specification applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units 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 This specification applies to the release of gaseous effluents from all reactors at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.3/4.11.2.2 DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept"as low as is reasonably achievable".

The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses 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. The ODCM equations provided for determining the air doses at the site boundary are based upon the historical average atmospheric conditions.

3/4.11.2.3 DOSE, RADIOIODINES, RADIOACTIVE

MATERIALS

IN PARTICULATE

FORM AND RADIONUCLIDES

OTHER THAN NOBLE GASES This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Opera-tion are the guides set forth in Section II.C of Appendix I. The ACTION state-ments provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable".

The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcula-tional procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses 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 The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the station operation.

This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.The initially specified monitoring program will be effective for at least the first three years of commercial operation.

Following this period, program changes may be initiated based on operational experience.

The detection capabilities required by Table 4.12-1 are state-of-the-art for routine environmental measurements in industrial laboratories.

The LLD's for drinking water meet the requirements of 40 CFR 141.3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey information from the door-to-door, aerial or consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.To determine this minimum garden size, the following assumptions were used, 1)that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/square meter.3/4.12.3 INTERLABORATORY

COMPARISON

PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of-radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.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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NRC Generic Letter 1979-30

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NRC Generic Letter 1979-30

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