ML17317B057

Radiological Effluent Tech Specs for DPR-58 & DPR-74

ML17317B057
Person / Time
Site:	Cook
Issue date:	03/27/1979
From:	INDIANA MICHIGAN POWER CO.
To:
Shared Package
ML17317B056	List: ML17317B054 ML17317B057 ML17317B059
References
NUDOCS 7904050205
Download: ML17317B057 (94)

Text

EFFLUENT TECHNICAL "SPECIFICATIONS INDIANA 8( l1ICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT UNIT NOS. 1 BRIDGMAN, MICHIGAN 00cket tea->r~

Control S > to SS~~<~~

at Ihcumenb RM Docket Nos. 50-315 and 50-316 License Nos.DPR-58 and DPR-74 g g04050 +~

The attached Effluent Technical Specification for the Oonald C. Cook Nuclear Plant is based on existing Unit 1 Appendix A Paragraphs. For Unit 2 Technical Specifications, use the following cross-reference list.

Unit 1 Corresponding Unit 2 Technical Specification Technical Specification Number Number 3/4.3.3.8 3/4.3.3.9

~

Table 3.3-11 Table 3.3-12 Table 4.3-11 Table 4.3-12 3/4.3.3.9 3/4.3.3.10 Table 3.3-12 Table 4.3-13 Table 4.3-12 Table 4 '-13 Bases 3/4.3.3.8 Bases 3/4.3.3.9 Bases 3/4.3.3.9 Bases 3/4.3.3.10

1.0 OEFINITIONS CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that is responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alrm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

The CHANNEL CALIBPATION may be performed by any series of sequential, over-lapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.1O A CHAl)ilEL CHECK shall be -the qualitative assessment of channel be-havior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrumentation channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TESTS 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.

SOURCE CHECK 1.27 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

1.0 DEFINITIONS (Continued)

OFFSITE DOSE CALCULATION MANUAL ODCM) 1.28 An OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be a. manual contain-ing the methodology and parameters to be used in the calculation of off-site doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm/trip setpoints. Requirements of the ODCM are provided in Specification 6.15.

GASEOUS RAD3IASTE TREATMENT SYSTEM 1.29 A GASEOUS RAD'llASTE TREATMENT SYSTEM is any system designed and in-stalled 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 EXHAUSE TREATMENT SYSTEM 1.30 A VENTILATION EXHAUST. TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in parti-culate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers 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.

TABLE 1.2 FRE UENCY NOTATION NOTATION FRHRUENCY At least once per 12 hours.

At least once per 24 hours.

At least once per 7 days.

At least once per 31 days.

At least once per 92 days.

SA At least once per 184 days.

At least once per 18 months.

S/U Prior to each reactor startup.

Completed Prior to each release.

N.A. Not applicable.

INSTRUMENTATION RAOIOACTIVE LI UIO EFFLUENT INSTRUMENTATION LIMITING COND ITIGN FOR OPERATION 3.3.3.8 The radioactive liquid effluent montoring "instrumentation channels shown in Table 3.3-11 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded.

APPLICABILITY: As shown in Table 3.3-11.

ACTION'.

With a radioactive liquid effluent monitoring instirumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of 3.11.1.1 are met, imme-diately suspend 'the release of radioactive liquid effluents monitored, by the affected channel or declare the channels in-operable.

b. With one or more radioactive liquid effluent monitoring in-strumentation channels inoperable, take the applicable ACTION shown in Table 3.3-11.

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

SURVEILLANCE RE UIREMENTS 4.3.3.8.1 The setpoints shall be determined in accordance with methodology as described in the OOCM and shall be recorded in the Technical Oata Book.

4.3.3.8.2 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 during the MOOES and at the frequencies shown in Table 4.3-11.

4.3.3.8.3 Records - Auditable records shall be maintained, in accordance (ith aaetho~do ogyin the 00Civi, of all radioactive liquid effluent monitoring instrumentation alarm/trip setpoints, Setpoints and setpoint calculations shall be available for review to ensure that the limits of Specification 3.11.1.1 are met.

TABLE 3.3-11 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION HI NI MUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Liquid Radwaste Effluent Line (1) At times of release 23

b. Steam Generator Blowdown (1) At times of release Line

c. Steam Gen. Blowdown Trt. Effluent (1) During primary to secondary leakage

2. Gross Radioactivity Monitors Not Providing Automatic Termination of Release

a. Service Water System Effluent Line At all times

3. Flow Rate Measurement Devices*

a. Liquid Radwaste Effluent Line At times of release 26 (1)

b. Discharge Pipes* (1) At all times

c. Steam Generator Blowdown Trt. (1) During primary to secondary 26 Effluent leakage

• Pump curves, and valve settings may be utilized to estimate flow; in such cases, action statement 26 is not required.

TMLE 3.3- ntinued)

RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION t1IN I NUM CIIANNELS INSTRUMENT OPERABLE APPL I CAB ILIT Y ACTION

4. Continuous Composite Samplers

a. Turbine 8uilding Sumps Effluent (I) At all times 25 Line

TABLE 3.3-11 (Continued)

TABLE NOTATION ACTION 23 >lith, the number of channels OPERABLE less than required by the Minimum Channels 8PERABLE requirement, effluent releases may be resumed for up to 14 days, provided that prior to initiating a release:

l. At least two independent amples are analyzed in accordance with Specification 4.11.1.1.3, and;

2. At least two technically qualified members of the Facility Staff independently verify the discharge valving.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 24 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 14 days provided grab samples are analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10-7 uCi/gram;

l. At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the specific activity of the secondary coolant is >0.01 pCi/gram OOSE EQUIVALENT I-131.

2. At lease once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is < 0.01 pCi/gram OOSE E(UIVALENT I-131.

ACTION 25 Mith the numbers of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided that at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-" pCi/ml.

ACTION 26 lilith the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for'p to 14 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.

TABLE 4.3-1l RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CIIECK CHECK CALIBRATION TEST

1. Gross Beta or Gamma Radioactivity Monitors Providing Alarm and Auto-matic Isolation.

a. Liquid Radwaste Effluents Line D* C(l)

b. Steam Generator Blowdown Effluent D* a(l)

Line Steam Generator Blowdown Trt.

Effluent

2. Gross Beta or Gama@ Radioactivity Monitors Providing Alarm But Nqt Providing Automatic Isolation L3)

a. Service Mater System Effluent Line D* M O(2)

3. Continuous Composite Samplers

a. Turbine Building Sumps Effluent N/A Line

• During releases via this pathway.

TABLE 4.3-1 ntinued)

RADIOACTIVE LI(}UID EFFLUENT MONITORING IN RUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

4. Flow Rate Monitors

a. Liquid Radwaste Effluent (q) N.A.

b. Steam Generator Blowdown Treatment Line D(<) N.A.

TABLE 4.3-11 (Continued)

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

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

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

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

(3) This requirement is applicable only to systems where the service water system or component cooling water system is discharged to an effluent stream.

(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.

INSTRUMENTATION RADIOACTIVE GASEOUS PROCESS AND EFFLUENT 'MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive gaseous process and effluent monitoring instru-mentation 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.F 1 are not exceeded.

APPLICABILITY: As shown in Table 3.3-12.

ACTION:

a e With a radioactive gaseous process or effluent monitoring in-trumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of 3.11.2.1 are met, declare the channel inoperable.

With one or more radioactive gaseous process or effluent monitoring instrumentation channels inoperable, take the applicable ACTION shown in Table 3.3-12.

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

SURVEILLANCE RE UIREMENTS 4 '.3.9.1 The setpoints shall be determined in accordance with methodology as described in the 00CM and shall be recorded in the Technical Data Book.

4.3.3.9.2 Each radioactive gaseous process or effluent monitoring instrumen-tation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, AND CHANNEL FUNCTIONAL TEST opera-tions during the MODES and at the frequencies shown in Table 4.3-12.

4.3.3.9.3 Auditable records shall be maintained of the calculhtions made, in accordance with methodology in the OCDM, of all radioactive process and effluent monitoring instrumentation alarm/trip setpoints. Setpoints and setpoint calculations shall be available for review to ensure that the limits of Specification 3.11.2.1 are met.

TABLE 3.3-'12 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABI E APPL I CAB ILI TY PARAMETER ACTION

1. l(aste Gas Holdup System Explosive Gas Monitoring System.

a. Ilydrogen Monitor (1) X Hydrogen 31

b. Oxygen Monitor (2) I Oxygen 29

2. Condenser Evacuation System

a. Noble Gas Activity Radi oacti vi ty Rate Measurement

b. Effluent System Flow Rate Measuring Device System Flow Rate Measurement 27

3. Auxiliary Building Ventilation System

a. Noble Gas Activity Radioactivity Rate Measurement 28 Moni tor

b. Iodine Sampler Cartridge Verify presence of cartridge 30

c. Particulate Sampler Filter (1) Verify presence of filter 30

d. Effluent System Flow (1) System Flow Rate Measurement 27 Rate Measuring Device

e. Sampler Flow Rate Sampler Flow Rate Measurement 27 Measuring Device

• During release via this pathway

• • During waste gas holdup system operation (treatment for primary system gases).

TABLE 3.3-12 (Continued)

TABLE NOTATION ACTION 27 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, for more than 28 days with the flow rate being estimated at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, pre-pare a Special Report to the Commission pursuant to Specifica-tion 6 '.2. within the next 10 days outlining the cause of the malfunction and plans for restoring the channel to OPERABLE status.

ACTION 28 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement for more than 28 days with grab samples being taken at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, prepare a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and plans for restoring the channel to OPERABLE status.

ACTION 29 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement for more than 14 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2. With both channels INOPERABLE for more than 14 days prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2, within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel(s) to operable status.

ACTION 30 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement for more than 28 days with samples being collected for periods on the order of seven (7) days and analyzed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the end of sample collection, prepare 5 Special Report to the Commission pursuant to Specification 6.9.2, within the next 10 days outlining the cause of the malfunction and the plans for restoring the channels to OPERABLE status.

ACTION 31 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement for more than 14 days with grab samples being taken and analyzed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, prepare a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and plans for restoring the channel to OPERABLE status.

TABLE 4. 3 12 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

l. Waste Gas Holdup System Explosive Gas Monitoring System

a. Hydrogen Monitor N.A. O (3)

b. Oxygen Monitor N.A. R (3)

c. Oxygen Monitor (alternate) N.A. O (3)

2. Condenser Evacuation System

a. Noble Gas Activity Monitor D

• a(2)

b. System Effluent Flow Rate N.A.

3. Auxiliary Building Ventilation System

a. Noble Gas Activity Monitor a(1)

b. Iodine Sampler N.A. N.AD N.A.'.A.

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

d. System Effluent Flow Rate Measurement Device D

• N.A.

e. Sampler Flow Rate Measurement Device N.A.

• During releases via this pathway

"*During waste gas holdup system operation (treatment for primary system offgases)

0 TABLE 4.3.12 TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the following condition exist:

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

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

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

(3) The CHANNEL CALISRATION shall include the use of standard gas samples containing: Anominal

1. High purity nitrogen (zero gas)

2. '% oxygen, 80/ hydrogen and the balance nitrogen (span aas)

3. Air-(span gas)

!3/4.11 RADIOACTIVE EFFLUENTS

'3/4.11 ~ 1 LI UID EFFLUENTS

'CONCENTRATION LIHITING CONDITION FOR OPERATION 3.11.1.1 The concentration of readioactive material released at anytime from the site to unrestricted areas (see Figure 3.11-1) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix 8, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.

For dissolved or entrained noble gases, the concentration shall be limited to 2 X 10-4 pCi/ml total activity.

APPLICABILITY: At all times ACTION:

With the concentration of radioactive material released from the site to unrestricted areas exceeding the above limits, immediately restore con-f centrati on wi thin the above limits and provide prompt noti i cati on to the Commission pursuant to Specification 6.9.1.12.

SURVEILLANCE RE(}UIREHENTS 4.11.1.1.1 The concentration of radioactive material at any time in liquid effluents releasedfrom the site shall be continuously monitored in accordance with Table 3.3-11.

4. 11. 1.1.2 The liquid effluent continuous monitors having provisions for automatic termination of liquid releases, as listed in Table 3.3-11, shall be used to limit the concentration of radioactive material released at any time from the site to unrestricted areas to the values given in Speci-fication 3.11.1.1.

4.11.1.1.3 The radioactivity content of each batch of radioactive liquid caste to be discharged shall be determined prior to release by sampling nd analysis in accordance wi thTable 4. 11-1. The results of pre-release analyses shall be used with the calculational methods in the ODCH to assure that the concentration at the point of release is limited to the values in pecification 3.11.1.1.

. 11.1,1.4 Release analyses of samples from batch release shall e performed in accordance with Table 4.11.-1. The results of the post-elease analyses shall be used wi th the calculational methods in the ODCH to ssure that the concentrations at the point of release are limited to he values in Specification 3.11.1,1.

SURVEILLANCE RE UIRE/1ENTS (Continued 4.11.1.1.5 The radioactivity concentration of liquids discharged from continuous relase 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 OOCM to assure that the concentrations at the point of release are limited to the values in Specification 3.11.1.1.

6.11.1.1.6 ~Re orts. The seniannual Radioactive Effluent Release Report shall include the information specified in Specification 6.9.1.9.

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TABLE 4.11-1(Continued}

TABLE NOTATION

a. The lower limit of detection (LLD) is defined in Table Notation a. of Table 4.12-1 of Specification 4.12.1.1.

b, For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radio-nuclides in concentrations near the LLD. Under these circumstances, the LLD may be increased inversely proportionally to the magnitude of the gamma yield (i.e. , 5 X 10 / I, where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for specific radionuclide, be greater than 10Ã of the HPC value specified in 10 CFR 20, Appendix, B, Table II, Column 2.

c. 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 repre-sentative of the liquids released.

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

e. A continuous release is the discharge of liquid wastes of a non-discrete 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 will apply are exlusively in the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, tlo-99, Cs-134, Cs-137, Ce-141, and Ce-144.

This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances reult in LLD's higher than required, the reasons shall be documented in the semiannual Radioactive Effluent Release Report.

RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to an individual from radioactive material in liquid effluents released to unrestricted areas (see Figure 3.11-1) shall be limited:

v

a. During any calendar quarter to < 1.5 mrem to the total body and to < 5 mrem to any organ, and

b. During any calendar year to < 3 mrem to the total body and 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, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceed-ing 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 average 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. (Applicable only is taken from the receiving water body.)

if drinking water supply

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

SURVEILLANCE RE UIREMENTS 4.11.1.2 ~ 1 Dose Calculations . Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calcu-lation Manual (ODCM) at least once per 31 days.

4.11.1.2.2. ~Re orts. The semiannual Radioactive Effluent Release Report shall include the information specified in Specification 6.9.1.9.

Figure 3.1l-l 00J-I It I II I II II Ii II 9

I'II I Ill II I

( I III I

) II I 4 0 4 0 Planl Nor th 0 0 goth+

0 Plant Discharge(unrestricted area) o Plant Intake

RADIOACTIVE EFFLUENTS LIQUID WASTE TREATHENT LIHITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be OPERABLE. The system shall be used to reduce the radioactive materials in liquid wastes prior totheir discharge when the projected dose due to liquid effluent releases to unrestricted areas (see Figure 3.11-1) when averaged over 31 days would exceed 0.25 mrem to the total body or 0.8 mrem to any organ.

APPLICABiLITY At all times.

ACTION:

a ~ With radioacti ve liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:

l. Identification of equipment or subsystems not OPERABLE and

~

the reason for inoperability.

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

3. Summa'ry description of action(s) taken to prevent a recurrence.

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable; URVEILLANCE RE UIREHENTS 4.11.1.3.1 Doses due to liquid release to unrestricted areas shall be projected at least once per 31 days, w'henever liquid release are being made ithout being processed by the liquid radwaste treatment system.

4.11.1.3.2 The liquid radwaste system shall be demonstrated OPERABLE at lease once per 92 days unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 days.

't RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate, at any time, in the unrestricted areas (see Page 3/4 12-6)due to radioactive materials released in gaseous effluents from the site shall be limited to the following values.

a. The dose rate limit for noble gases shall be <500 mrem/yr to the total body and < 3000 mrem/yr to the skin, and

b. The dose rate limit for all radioiodines and for all radioactive materials in particulate form and radionuclides other than noble gases with half lives greater than 8 days shall be < 1500 mrem/yr to any or gan.

APPLICABILITY: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, immediately decrease the release rate to comply with the limit(s) given in Specification 3.11.2.1 and provide prompt notification to the Commission pursuant to Specifica-tion 6.9.1.12.

SURVEILLANCE RE UIREi4IENTS 4.11.2.1.1 The release rate, at any time, of noble gases in gaseous effluents shall be controlled by the offsite dose rate as established above in Specification 3.11.2.1.

4.11.2.1.2 The noble gas effluent continuous moni tors having provisions for the automatic termination of gaseous release, as listed in Table 3.3-12, shall be used to limit offsite doses within the values established in Specification 3.11.2.1 when monitor setpoint values are exceeded.

4. 11.2.1.3 The release rate of radioacti ve materials, other than noble gases, in gaseous effluents shall be determined by obtaining representative samples and performing analyses in accordance with the sampling and analysis program, specified in Table 4.'il-2.

SURVEILLANCE REOUIREHENTS

'4.11.2.1.4 The dose rate in unrestricted areas, due to radioactive materials other than noble gases released in gaseous effluents, shall be determined to be within the required limits by using the results of the sampling and analysis program, specified in Table 4.11-2, in performing the calculations of dose rate in unrestricted areas.

4.11.2.1.5 ~Re orts The semiannual Radioactive Effluent Release Report shall include the information specified in Specification 6.9.1.9.

, RADIOACTIVE EFFLUENTS DOSE, NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose in unrestricted areas (see Page 3/4 12-6)due to noble gases released in gaseous effluents shall be limited to the following:

a. During any calendar quarter, to < 5 mrad for camma radiation and < 10 mrad for beta radiation;

b. During any calendar year, to < 10 mrad for gamma radiation and

< 20 mrad for beta radiation.

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits. oreoare and submit to the Commission within 30 davs. oursuant to Soeci-fication 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 average 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 RE UIREMENTS 4.11.2.2.1 Dose Caculations Cumulative dose contributions for the total time period shall be determined in accordance with the Offsite Dose Calculation Manucal (ODCM) at least once every 31 days.

4.11.2:2.2 ~Re orts The Semiannual Radioactive Effiuent Re1ease Report shall include the information specified in Specification 6.9.1.9.

RADIOACTIVE EFFLUENTS DOSE, RADIOIODINES, RADIOACTIVE t1ATERIAL IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES LIi~1ITING CONDITION FOR OPERATION 3.11.2.3 The dose to an individual from radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days in gaseous effluents released to unrestricted areas (see Pa9e 3/14 12-6)hall be limited to the following:

a. During any calendar quarter to < 7.5 mrem to any organ; b.. During any calendar year to < 15 mrem to any organ.

APPLICABILITY: At all times ACTION:

a. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than nobles gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission with-in 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 radioiodines, radioactive materials in particu-late form, and radionuclides other than noble 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 average 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.

id~ii SURVEILLANCE REQUIREMENTS 4.11.2.3.1 Dose Calculations Cumulative dose contributions for the total p di 4 ih h OC every 31 days.

4.11.2.3.2 ~Re orts The semiannual Radioactive Effluent Release Report shall include the information specified in Specification 6.9.1.9.

TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Mlmlmum Lower Limit of Gaseous Release Type Frequency Analysis Type of Detection (LLD) t Frequency Activity Analysis > Ci ml a P P A. Waste Gas Storage Each Tank Each Tank Princi al Gamma Emitters. 1 X 10-4b Grab Sam le H-3 1X10 B. Containment Purge P P Principal Gamma Emitters" 1X10 Each Purge Each Purgec rab H-3 1 X 10-6 C.

Air Ejector Mc Principal Gamma Emitters f 1 X 10 4b Grab Sample H-3 1 X 10-6 I -131 1 X 10-12 D. Auxi 1 i ary Building Continuous Charcoal Vent Sample I-133 1X10 Continuous Particulate Principal Gamma Emitters f

Sample (I-131, Others) 1 xlo M

Continuous e

Composite Gross alpha 1X10 Par ti cul ate Sample Continuous e

Composite Sr-89, Sr-90 1 xlo Particulate Sample

TABLE 4.11-2 Conti nued)

TABLE NOTATION The lower limit of detec'tion (LLD) is defined in Table Notation a. of Table 4.12-1 of Specification 4.12.1.1.

For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in conentrations near the LLD. Under these circum-stances, the LLD may be increased inversely proportionally to the magnitude of the gamma yield (i.e., 1 X 10 4 /I, where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for speci'fic radionuclide, be greater than 10Ã of the MPC value specified in 10 CFR 20, Appendix B, Table II, Column l.

Analyses shall also be performed following shutdown, startup, or similar operational occurrence which has altered the mixture of radionuclides as indicated by RCS analysis.

Analyses shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for 7 days following each shutdown, startup or similar operational occurrence which lead to significant increases or decreases in radioiodine in the Reactor Coolant System. Samplers shall also be changed and analy'ze'd 'at th'e intervals in Specifications 3.11.2.1 and.2.11.2.3. When samples collected for 24 hourt are anlayzed, the corresponding LLD's may be increased by a factor of 10.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.3.3 and 3.11,3.3.

The principal gamma emitters for which the LLD specification will apply are exclusively 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 -missions. 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 should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the semi-annual effluent report.

RAOIOACTIVE EFFLUENTS

'GASEOUS RADNASTE TREATMENT LIMITING CONOITION FOR OPERATION 3.11.2.4 The gaseous radwaste treatment system and the venti'lation exhaust treatment system shall be OPERABLE. 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 to unrestricted areas (see Page 3/4 12-6) when averaged over 31 days would exceed 0.8 mrad for gamma radiation and 1 6 mrad for beta radiation and ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their dicharge when the projected doses due to gaseous effluent releases to unrestricted areas (see Page 3/4 12-. 6) when averaged over 31 days would exceed 1.25 mrem to any organ, APPLICABILITY: At all times.

ACTION:

a. With gaseous waste being discharged for more than 31 days without treatment and in excess of the. above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:

l. Identification of equipment of subsystems not OPERABLE and the reason for inoperability.

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

3~ Summary description of action(s) taken to prevent a re-currence.

b. The provisions of Specification 3.0.3 and 3.0.4 are not appl i cab 1 e.

SURVEILLANCE RE UIREMENTS 4.11.2.4.1 Ooses due to gaseous releases to unrestricted areas shall be projected at least once per 31 days, whenever the gaseous waste treatment system or ventilation exhaust treatment system is not in operation.

4.11.2.4.2 The appropriate systems shall be demostrated OPERABLE at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION I

3.11.2.5 The dose or dose commitment to a real individual 'from all

-'ranium fuel cycle sources is limited to < 25 mrem to the total body or any organ (except the thyroid, which is limited to < 75 mrem) over a

'eriod of 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 Specifications 3.11.1.2.a, 3.11.1.2.b, 3.11.3.3.a F 11.2.2.b, 3.11 '.3.a, or 3.11..2.3.b, prepare and submit a Special Report to the Commission pursuant to Speciication 6 .9 .2 and limit the subsequent releases such that the dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to < 25 mrem to the total body or any organ (except thyroid, which is limited to < 75 mrem); over 12 consecutive months. This Special Report shall include an analysis i which demonstrates that radiation exposures"to all real individuals from all 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 exceeds the 40 CFR Part 190 Standard.

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

SURVEILLANCE RE UIREMENTS 4.11.2.5.1 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with.Specifi-cations 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, and 3.11.2.3.b, and in accordance with the Offsite Dose Calculation Manual (ODCM).

4.11.2.5.2 ~Re orts Specia1 Reports sbal1 be subiaitted as required under Specification 3.11.2.5.a.

RAOIOACTIVE EFFLUENTS

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

(

LIMITING CONPITON FOR OPERATIO 3.11.2.6 The concentration of oxygen in the waste gas holdup system shall be limited to < 2% by volume if the hydrogen'in the system is

> 4% by volume.

APPLICABILITY: At al 1 times.

ACTION:

a. With the concentration of oxygen in the waste gas holdup system

> 2% by volume but < 4% by volume and containing > 4% hydrogen, restore the concentration of oxygen to< 2/or reduce the hydrogen concentration to < 4% within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b. With the concentration of oxygen in the waste gas holdup system or tank >4% by volume and>4% hydrogen by volume immediately suspend all addi tions of waste gases to the system or tank and reduce the concentration of .oxygen to < 2% or the concentration of hydrogen to

< 4% within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in the system or tank.

c. /he provisons of Specification 3.0.3 and 3.0.4 are not applicable.

URVEILLANCE RE UIREMENTS

.11,2.6 The concentrations of oxygen in the waste gas holdup system shall be determined to within the above limits. by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitors required OPERABLE by Table 3.3-12 of Specification 3.3.3.9.

RADIOACTIVE EFFLUENTS GAS STORAGE TANKS Lli4lITING CONDITION FOR OPERATION 3.11.2.7 The quantity of radioactivity contained in each gas storage tank shall be limited to < 438,000 curies noble gases (considered as Xe-133).

APPLICABILITY: At al 1 times ACTION:

a. With the quantity of'adioactive 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> either reduce the tank contents to within the limit or provide prompt notification to the Commission pursuant to Specification 6.9.1.12. The written followup report shall include a description of activities planned and/or taken to reduce the tank contents to within the above limit.

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

SURVEILLANCE RE UIREf1ENTS 4 ~ 11,2 7 The quantity of radioactive material contained in each gas

~

storage tank sha',1 be determined to be within the above limit at least once per 4 days by analysis of the Reactor Coolant System noble gases.

3/4.12 RAOIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONOITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be con-ducted 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. (Oeviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavail-ability, or to malfunction of automatic sampling equipment. If the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.)

b. With the level of radioactivity in an environmental sampling medium at one or more of the locations specified in Table 3.12-1 exceeding the limits of Table 6 '.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 Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table 6.9-2 to be exceeded. 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.

0 c ~ With milk or fresh leafy vegetable samples unavailable from any of the sample locations required by Table 3.12-1, prepare and submit to the Commission within 30 days, pursuant to Specifi-cation 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 un-available may then be deleted from Table 3.12-1 provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations, if available.

d. The provisions of Specifications 3.0.3 and 3.0.4 are not appl i cab 1 e.

SURYEILLANCE RE UIRB1ENTS 14.12.1.1 The radiological environmental monitoring samples shal 1 be collected pursuant to Table 3.12-1 from the locations shown on Figure 3.12-1 and shall be analyzed pursuant to the requirements of Tables 3.12-1 and 4.12-1.

4.12.1.2 ~Re orts The results of analyses performed on the radio logical environmental monitoring samples shall be summarized in th;.

annual Radiological Environmental Operating Report, pursuant to Specification 6.9.1.6.

TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Type and Frequency

/~s Sam le Locations** Collection Fre uenc of Anal sis

1. AIRBORNE A. Radioiodine Al-A6 Continuous operation of Radioiodine canister.

and Particulates NBF, SBN, DOW, COL sampler" with sample col- Analyze at least once per lection as required by 7 days for I-131 dust loading but at least once per 7 days. Particulate sampler.

Analyze for gross beta radioactivity >24 hours following filter change.

Perform gaoma isotopic analysis on each sample when gross beta activity is

> 10 times the mean of control sample. Perform gamm<

isotopic analysis on com-posite (by location) sample at least once per 92 days.

2. DIRECT RADIATION Al-A6 At least once per 92 days Gamma dose. At least once NBF, SBN, DON, COL

• per .92 days.

• Direct radiation TLD badges located with each air sampler

• • Sample locations are shown on Figure 3.12-1.

TABLE 3.12-1 continued RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Type and Frequency and/or Sam le Collection Fre uenc of Anal sis

3. WATERBORNE

a. Surface Composite

• sample collected Gamma isotopic analysis over a period of < 31 days. of each composite sample by location. Tritium analy-sis of composite sample at least once per 92 days.

b. Ground Wl -W7 At least once per 92 days Gamma'isotopic and tritium analyses of each sample.

c. Drinking St. Joseph Composite* sample collected Gross beta and gamma Lake Township over a period of < 31 days. isotopic analysis of each New Buffalo composite sample. Tritium analysis of composite sample at least once per 92 days.

d. Sediment from L2, L3 2/year Gamma isotopic analysis Shoreline of each sample.

4~ INGESTION '.

Milk Stevensvi lie At least once per 15 days Gamma isotopic and I-131 when animals are on pasture; analysis of each sample.

Bridgman at least once per 31 days Galien at other times.

Dowagiac South Bend

• Composite samples shall be collected by collecting an aliquot at intervals not exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

• • Sample Locations are shown on Figure 3.12-1.

TABLE 3.12-1 Continued RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Type and Frequency

~d/ ~Sam le Locations ** Collection Fre uenc of Anal sis

4. INGESTION (cont.)

b. Fish 2/year Gamma isotopic analysis Plant site on edible portions.

Off-Site

c. Food Products At time of harvest. One sample Gaoea isotopic analysis Plant Si te of each of the following classes on edible portion.

Off-Site (approx. 20 miles)of food products:

~

1. Grapes Plant Site At time of harvest. One Gamma isotopic analysis.

sample of broad leaf vegetation.

• At least 5 miles from plant centerline.

• • Sample locations are shown on Figure 3.12-1.

Figure 3,'i2-1A

- Air, Precipitation, TLD Stations

- Lake Mater Sample Stati ons

- Milk Sample Stations 20 MILES ~, I196 llati.i Col om 194 BENTON HARBOR ST. JOSEPH M140 Stevensvil e 33 D. C. COOK PLANT Q Eau

,/ Bridgma laire Berrien M Springs I94 Ni les L

New Buffalo US 12 A

MECHIGAH 17lDTFiiA MICHIGAN CITY New Carlisle US20 r

US20 IND 2 'SOTH BlD Scale of Miles 10 20

TRUE PLANT UNRESTRICTED AREA NORTH NORTH

~ NF

~

PROPERTY LI Ill Al ROADS

)

H2 Ll I I]4 345 KV YARD (l!.

f ~E-V~PLANT q 765 KV SHORE LINE L2 r

N.S PLANT C YARD CHESAPEAKE a OHIO R.R.

2,000 FOOT RADIUS INTERSTATE 94 SCALE I,GOO 2,000 3,000 4,000 FEET A - Air, Preci pi tati on, TLD Stati ons W - tlell Water Sample Stations I. - Lake Water Sample Stations M - Milk Sample Stations

TABLE 4.12-1 MAXIMUM VALUES FOR Tllf LOWER LIMITS OF DETECTION (LLD)

Water Airborne Particulate Analysis (pCi/1) or Gas Fl sb Mi 1 k Food Products Sediment (pCi/m3) (pCi/kg,wet) (pCi/1) (pCi/kg, wet) (pCi/kg,dry) ross beta 4b 1X102 3 2000 (1000 )

ll 54Mn 15 130 59FFe 30 58,60CCo 15 130 Zn 30 260 Zr-Nb 1311 7 X 10"2 60 134,137 15(10b),18 1X102 130 15 80 150 Cs 140 15 15 Ba-La

TABLE 4.12-1 (Continued)

TABLE NOTATfON.

a - The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95/ probability with 55 probability of falsely concluding that a blank observation represents a "real" signal.

For a particualr measurement system (which may include radio-chemical separation):

LLD = 4.66 s E ~ V 2. 22 ~ Y exp(-Rat) where LLD is -the lower limit of detection as defined above(as pCi per unit mass or volume) s is the standard deviation of the background counting rate ok of 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 tansformation per minute per curie, is the fractional radiochemical yield (when applicable)

A i s the radi oacti ve decay cons tant for the parti cul ar radionuclide at is the elapsed time between sample collection (or end of the sample collection period) and time of counting.

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 predi cted 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 mil,k samples).

TABLE 4.12-1 Continued)

TABLE NOTATION Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, .unavoidably small sample sizes, the presence of interferring nuclides, or other uncontrollable cir-cumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

LLD for drinking water.

LLD for leafy vegetables.

l 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 loca-tiori of the nearest milk animal, the nearest residence and the nearest garden* of greater than 500 square feet producino fresh leafy vegetables in each of the 9 land covering meterological sectors within a distance of five 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, pre-pare 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 5 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, 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, if possible.

The sampling 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 RE UIREMENTS 4.12.2.1 The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1, by door-to-door survey, aerial survey, or by consulting local agriculture authorities.

4.12.2.2 ~Re orts The results of the land use census shall be included in the Annual Radi ol ogi ca 1 Environmental Operating Report.

Broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest 0/g in lieu of the garden census.

INSTRUMENTATION

.BASES

'RAOIOACTIVE LI UIO EFFLUENT INSTRUMENTATION 3/4.3.3.8 The radioactive liquid effluent instrumention is provided to monitor and control, as applicable, the releases of radioa'ctive materials in liquid effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the OOCM 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 Oesign Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.9 RAOIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION 3/4.3.3.9 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. The alarm/trip setpoints for these intruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumen-tation also includes provisions for monitori ng the concentrations of potentially explosive gas mixtures in the waste gas holdup'ystem.

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

3/4.11 RADIOACTIVE EFFLUENTS BASES.

3/4.11.1 LI UID EFFLUENTS 3/4.11.1.1 CONCENTRATION. This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix 8, Table II. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures within (1) the Section II.A design objec-tives of Appendix I, 10 CFR Part'0, to an individual and (2) the limits of 10 CFR Part 20,106(e) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotooe and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protec-tion ( 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 flexi6ility 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 effluents 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 is to be shown by cal-culational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially under-estimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents will be con-sistent 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. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113.

PHR-STS-1

RADIOACTiVE EFFLUENTS BASES This specification applies to the release of liquid effluents from

,each reactor at the site. The liquid effluents from the shared system are proportioned among the units sharing the system.

3/4.11.1.3 LIAUID HASTE 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

';requirements that the appropriate portions of this sytem be used when speci-

.fied provides assurance that the releases of radioactive materials in liquid

'.effluents will be kept "as low as is reasonably achievable." This speci-fication implements the requirements of 10 CFR Part 50.36a, General Design

,"Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section 11.D of Appendix A to 10 CFR Part 50. The specified limits governing the use

of appropriate portions of the liquid radwaste treatment system were speci-

fied as a suitable fraction of the guide set forth in Section 11.A'of

'-Appendix I, 10 CFR Part 50, for liquid effluents.

3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE. This specification is provided to ensure that the dose rate at anytime 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. 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, to annual average concentrations exceeding the limits specified in Appendix 8, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).

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 PHR-STS-1

RADIOACTIVE EFFLUENTS BASES 1

I'.

in the atmospheric diffusion factor above that for the site boundary.

. The specified release rate limits restrict, at all times, the corre-

sponding gamma and beta dose rates above background to an individual

- at or beyond the site boundary to < (500) mrem/year to the total body

or to < (3000) mrem/year to the skin. These release rate limits also

• 'estrict, at all times, the corresponding thyroid dose rate above back-ground to an infant via the cow-milk-infant pathway to < 1500 mrem/year for the nearest cow to the plant.

This specification applies to the release of gaseous effluents from all reactors at the site. 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 IY.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 assure that the releases of radio-active material in gaseous effluents will be kept "as low as is reasonable achievable". The Surveillance Requirement implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be 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 or Estimating Atmos-pheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-!later-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at the exclusion area boundary will be based upon the historical average atmospherical conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.

3/4.11.2.3 DOSE, RADIOIODINES, RADIOACTIVE MATERIAL IN PARTICULATE FORA 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 Conditions for Operation are the guides set forth in Section II.C of Appendix I.

i PWR-STS-1

RADIOACTIVE EFFLUENTS BASES The ACTION statements provide the required operating flexi-bility 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 gaseous effluents will be kept "as low as is reasonably achievalbe."

The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conform-.

ance 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 substantially under-estimated. The ODCH calculational methods approved by NRC for calculating the doses due to the actual release rates of the subject materials are re'quired to be consistent with the methodology provided in Regulatory Guide 1. 109, "Calculating of Annual Doses to Han from Routine Releases of Reactor Effluents for the'Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, "Revision I, October 1977 and Regulatory Guide 1.111, "Hethods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"

Revision 1, July 1077. These equations also provide for determinino the actual doses based upon the historical average atmospheric condi tions.

The release rate specifications for radioiodines, radioactive material 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 are examined in the development of these calculations are: 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 consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

3/4.11.2.4 GASEOUS WASTE TREATMENT The OPERABILITY of the gaseous radwaste treatment system and the venti la-tion exhaust treatment systems ensures that the systems wi 11 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 Appen-dix A to 10 CFR Part 50, and design objective Section IID 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 guide set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

RADIOACTIVE EFFLUENTS 3/4. 11.2. 5 00SE This specification is provided to meet the reporting requirements of 40 CFR 190.

3 4.11.2.6 EXPLOSIVE GAS HIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treat-ment system is maintained below the flammability limits of hydrogen and oxygen mixtures. Haintaining the concentration of hydrogen or oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Oesign Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.11.2.7 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 tanks contents, the resulting total body exposure to an individual at the nearest site boundary will not exceed o.5 rem.

This is consistent with Standard Review Plan 15.7,1, "4'aste Gas System Failure."

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 measur-able 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 fi rst three years of commerical operation. Following this period, program changes may be initiated based on operational experience.

The detection capabilities requi red by TAble 4.12 '1 are state-of-the-art for routine environmental measurements in indiustrial labora-tories. 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.

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 ere used, 1) that 20/ of the garden was used for growing broad lea vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/square meter.

.6. 0 AOHIN ISTRAT I VE CONTROLS 6.1. RESPONSIBILITY 6.1.1 The plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.

6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Fi gure 6.2-1.

FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a ~ Each on duty shift shall be composed of at least the minimum shift crew composi tion shown in Table 6 '-1.

b. At least one licensed Operator shall be in the control room when fuel is in the reactor.

c ~ At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactory trips.

An individual qualified in radiation protection procedures shall be on si te when fuel is in the reactor.

e. ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervi sed by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

A site Fire Brigade of at least 5 members shall be maintained onsite at all times. The fire Brigade shall not include 3 members of the minimum shift crew necessary for safe shutdown of the unit or any personnel required for other essential functions during a fire emergency.

LEGEND:

ADMINISTRATIVE PRESIDENT TECHNICAL SUPERVISION AEP, AEPSC, I S'i'ND OTHER AEP 5 FUNCTIONAL DtRECTION SUBSIDI AR I c S TECHNICAL LIASON EXECUTIVE VICE PREStDENT VICE PRFSIOENT Yice INDtANA oi MICHIGAN NDI*NA5 MICHIGAN Chairman FLECTRIC COMPANY LECTRIC COiVi?ANY ENGINEERING 8( INDIANA & MICHtGAN I S MPO;";ER CO. CONSTRUCTION AEPSC POWER COMPANY

. Senior VICE PRFSIDENT CQ Iigg pi 0n

'ASSISTANT Chief OTHER QUA LITY TO NUCLEAR ENGliNEER ENGINEERING ASSURANCE DIV)SIONS MANAGER EXEC. V. P. AEPSC AEPSC AEPSC AEPSC lwSDRC 1 I

J PI.ANT PLANT MANAGER QUALITY ASSUR ANCE DONALD C. COOK f4UCLEAR PLANT SUPER VISOR li DI NA AND MICHIGAN ELECTRIC COPs PAN Y "ND INDIANAAND MICHIGAN PO'i'I'ER COM?A" Y

<<<<Responsible for Fire Pro ection Proc,ram.

r IGURE 6.21 Offsite Orcanization far Faciti~ Management arri Technical Suppoa

, PLNT f'NHAG hat.nt<<STIIATIVE MhINTf NANCF. ~ arE<<htlotl ~ TF.CIINICAL SUPE<<VISOA Sui) C. Sunt. SupC.

SOL mao. surv. ~

STAFF SThfF OPEAATIOtls SOL OA SursnvlSAA TAhlt<<tin SI AFT coo<<atwh ron OPEIIATINQ ENQ.

SO I.

OPE IIATltla EtioltlEEA NUCI.EAA ENGIWEEA rLnfonuhwcE SUI'EAVISOA conr<

din CIOFI OL EtlaltlEEA KIIGINEEA upv. I'rot.Supu LEGEtta: UNIT P E n Fa AtkhttCE PEAFOAMRtlcE ItlSTAUIAENT naothTloN IIhttITcttAIICE PAOT ECT loll SGL - sEtllon orEnhTOA LICEIISE Sgttv. ENGltIEEA FNOIIIEEA 'lny. CI I O.IIS'T Supv. oL - opEAATan LlcENsE S ~ - IIEv supEAVIEOAY PEnsoNNEL EGUIPI IEIIT OPEAATon T E CI INIClhlls 'f ECIINICIANS AUXILIhnY E0UA'I.IEtlT opEAATOB FlGUBt. 6.2-2 Fttdtlty OTgsnizathn 'onald C. Cook - Unit No. 1 TABLE 6.2-1 MINIMUM SHIFT CREM COMPOSITION LICENSE CATEGORY APPLICABLE tdOOES 1,2,3 & 4 5 & 6 SOL OL NON-Licensed ~ Ooes not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS after the initial fuel loading.
• • Shared with D. C. Cook Unit 2 8 Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accomodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within. the minimum requirements of Table 6.2-1.
ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF UALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable position, except for the Radiation Protection Superviosr who shall meet or exceed the quali-fications of Regulatory Guide 1.8, Spetember 1975. 6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18. 1-1971 and Appendix "A" of 10 CFR Part 55. 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Plant Manager and shall meet or exceed the require-ments of Section 27 of the NFPA Code-1976. 6:5',REVIEW AND AUOIT 6.5.1 PLANT NUCLEAR SAFETY REVIEM COMMITTEE PNSRC FUNCTION 6.5.1.1 The PNSRC shall function to advise the Plant Manager on all matters related to nuclear safety. AOHI NI STRATI VE CONTROLS COMPOSITION 6.5.1.2 The PNSRC shall be composed of the: Chai rman: Plant Manager or designated alternate Member: Asst. Plant Manager Member: Operations Superintendent lleaiber: Technical Superintendent Member: Maintenance Superintendent Member: Control and Instrument Enoineer Member: Nuclear Engineer Member: Plant Chemical Supervisor Member: Performance Supervisor Engineer Hember: Plant Radiation Protection Supervisor ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PNSRC Chairman to serve on a temporary basis; however., no more than two alternates shall participate as voting members in PNSRC activities at any one time MEETING FRE UENCY 6.5.1.4 The PNSRC shall meet at least once per calendar month and as convened by the PNSRC Chairman or his designated alternate. UORUM 6.5.14 A quorum of the PNSRC shall consist of the Chairman or his designated alternate and four members including alternates. RESPONSIBILITIES 6 '.1.6 The PNSRC shall be responsible for:

a. Review of 1) all procedures required by Specification 6.8 and changes thereto as determined by the Plant Manager to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

AOMIN I STRATI VE CONTROLS c ~ Review of all proposed changes to Appendix "A" Technical Specifications.

d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Chairman of the NSORC.

Review of those REPORTABLE OCCURENCES requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Commission.

g. Review of facility operations to detect potential safety hazards.

h ~ Performance of special reviews, investigations or analyses and reports thereon as requested by the Chair-man of the NSDRC. Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSORC. Review of every unplanned release of radioactive material to the environs; evaluate the event; specify remedial action to prevent recurrenc and document the event description, evaluation, and corrective action and the disposi tion of the corrective action in the plant records. AUTHORITY 6.5.1.7 The PNSRC shall Recorrmend to the Plant Manager written approval or disapproval of items considered under 6.5.1.6 (a) through (d) above. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6 (a) through (e) above constitutes an unreviewed safety question. c ~ Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NSDRC of disagreement between the PNSRC and the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above. ADMINI STRATI VE CONTROLS RECORDS 6.5.1.8 The PNSRC shall maintain written minutes of each meeting and copies shall be provided to the Chairman of the NSDRC. 6.5.2. NUCLEAR SAFETY AND DESIGiV REVIEM COMMITTEE(NSDRC) F UNCT ION 6.5.2.1 The NSDRC shall function to provide independent review and audit of designated activities in the areas of:

a. nuclear power plant operations

b. nuclear engineering

c. chemistry and radiochemistry

d. metallurgy

e. instrumentation and control

f. radiologi.cal safety

g. mechanical and electrical engineering

h. quality assurance practices.

'KPll+ tlVI i!Il i J ~ W > vl+ ', SI i rf<l ~ ~ all d."I I iv'v4 COflPOS IT IO'I 6.5.2.2 The NSDRC shall be composed of the: Vice Chair.".,an Engineering and Cons"ruction Senior Vice President Construction Executive 'lice President indiana ~ l!ichigan Electric Companly 'lice President Elec;rica]:-ngineerina Vice Presid r t l',ecnanical Enaineerina Assistan. Vice ~resident ard Chief Civil Encineer Chi e f tluc l ear - na i reer (Cha i rr.:an) Chief Design Enai. eer Plart Iianag r, Donald C. Cook Plant Head Environmental =n "ineering Division Head, Nuclear Safety A Licensing Section (Secretarv) Al ternate: Executive Assistan to the Vice Chairman Enaineering 5 Construction Aiterrate: Assistant Division Head, Project Control <<nd Support Oivis;on Alternate: Executive Assistant to the Executive 'lice President f 5 ll Alternate Assistant Chief t'.echanical Enaineer. ~ 'A'1'tern'ate".'~'Ass'istani Chief Civil Engineer Alternate: Assistant Division Head, Nuclear =ngineering Division Alternate: Head, Electrical Plant Qesian Sectior. Alternate: Assistant Plant,'tanager, Donald C. Cook Plan Alternate: Senior Staff Engineer, =nvi ronmental Enaineering Division Alternate: Encineer, nuclear Safety h Llcensira Section Alternate: A=;"-SC I';anager Gf guality Assulance Alternate: Assistant Chief Electrical Enlgineer ALTERNA'ES 6.5.2.3 All alternate members shall be appointed in writing by the N ORC Chairman to serve on a temporary basis; however, no lmore than-two alternates shall participat as voting members in NSORC activit'es at any one time. CONSULTANTS 6.5.2.4 Consultants.shall be uti1ized as det rmined by the NSORC Oir ctor to provide exper advic to the NSORC,. NEETTHG F?EOUENCY 6.5.2.5 The NSORC shall Ile << at least once per calend r auart r dur'ng the initial year o" facility operatior. ollowina fuel loading and at least once per six months thereafter. UORUH 6.5.2.6 A quorum of NSORC shall consist of the Chairman or his designated alternate and at leas. 4 I'(SORC members ircluding alternates. No mor than a m.'nority of the quorum shall have line responsibility =or operation of the facili-y. ADt1INISTRATIVE CONTROLS REVIEW 6.5.2.7 The NSDRC shall review:

a. The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.50, 10 CFR, to'erify that such actions did not constitute an unreviewed safety questions.

b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety questions as defined in Section 50.59, 10 CFR

c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR

d. Proposed changes in Technical Specifications or licenses.

e. Violations of applicable statutes,codes, regulations orders, Technical Specifications, license requirements or of internal procedures or instructions having nuclear safety significance.

f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

g. REPORTABLE OCCURRENCES requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to 'the Commission.

h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

i. Reports and meetings minutes of the PNSRC.

ADMIN I STRATI VE CONTROLS AUDITS 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the NSDRC. These audits shall encompass. a 0 The conformance of facility operation to provisions contained <vithin the Technical Specifications and applicable license conditions at lease once per 12 months.

b. The performance, training and qualifications of the entire facility staff at least once per 12 months.

c ~ The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least oncer per 6 months. d, The performance of activities required by the guality Assurance Program to meet the criteria of Appendix "B" 10 CFR 50, at least oncer per 24 months.

e. The Facility Emergency Plan and implementing procedures at least once per 24 months.

The Facility Security Plan and implementing procedures at least oncer per 24 months.

g. Any other area of facility operation considered appropriate by the NSDRC.

The Facility Fire Protection Program and implementing procedures at least once per 24 months. An independent fire protection and loss prevention program inspection and audit shall be performed at least oncer per 12 months utilizing either qualified offsi te licensee per-sonnel or an outside fire protection firm. An inspection and audit of the fire protection and loss pre-vention program shall be performed by a qualified outside fire consultant at least once per 36 months. The radiological environmental monitoring program and the results thereof at least once per 12 months. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least oncer per 24 months. AotdINISTRATIVE CONTROLS AUTHORITY 6.5.2.9 The NSORC shall report to and advise the Vice Choirman Engineering and Construction AEPSC, and those areas of responsibility specified in sections 6.5.2.7 and 6.5.2.8. RECOROS 6.5.2.10 Records of NSDRC activities shall be prepared, approved and distributed as indicated below:

a. Minutes of each NSDRC meeting shall be prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC, within 14 days following each meeting.

b. Reports of review encompassed by Section 6.5.2.7 above, shall be prepared, approved and forward to the Vice Chairman, Engineering and Construction, AEPSC within 14 days following completion of the review.

cd Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice Chairman, Engineering and Construc-tion, AEPSC, and to the management positions responsible for the areas audited within 30 days after completion of the audit. I 6.6 REPORTABLE OCCURRENCE'CTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Commission shall be reviewed by the PNSRC and submitted to the NSDRC and the Chief Nuclear Engineer.

ADMINISTRATIVE CONTROLS

6. 7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The facility shall be placed in at least HOT STANOBY within on hour..

b. The Safety Limit violation shall be reported to the Commission and to the Chairman of the NSDRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. A Safety Limit Violation Report shall be prepared. The report shall be revi'ewed by the PNSRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shal'1 be submitted to the Commission, the Chairman of the NSDRC and the Chief Nuclear Engineer within 14 days of the violation.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and main-tained covering the activites referenced below: a, The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

b. The radiological environmental monitoring program.

c. Refueling operations.

d. Surveillance and test activities of safety related equipment

e. Security Plan implementation,

f. Emergency Plan implementation.

g. Fire Protection Program implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PNSRC and approved by the Plant Manager prior to ivplementation and reviewed periodically as set forth in administrative procedures. ADMI NISTRATI YE CONTROLS .8.3 Temporary Changes to procedures of 6.8.1 above may be made rovided:

a. The intent of the original procedure is not.altered.

b. The change is approved by two members of the plant management staff, at least one of who holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented, reviewed by the PNSRC and approved by the Plant Manager within 14 days of implementation.

.9 REPORTING RE UIREMENTS OUTINE REPORTS AND REPORTABLE OCCURRENCES .9.1 In additicn to the applicable reporting requirements of Title 10, ode of Federal Regulations, the following reports shall be submitted o the Director of the Regional Office of Inspection and Enforcement nless otherwise noted. TARTUP RFPORT .9.1.1 A summary report of plant startup and power escalation testing shall be submitted following ( 1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) instaIlation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significani ly altered the nuclear, thermal, or hydraulic perfor-mance of the plant. 6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shalI include a description o f the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any correcti ve actions that. were required to obtain satisfactory operation shall also be described. Any additional specific details required in license condi tions based on other commit-ments shall be included in this report. ADMINISTRATIVE CONTROLS 6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months follow-ing initial criticality, whichever is earliest. If the Startup Re'port does not cover all three events (i.e., initial criticalitj, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submi tted at least every three months unti 1 all three events have been completed. ANNUAL REPORTS 6 '.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following .initial criticality. 6.9.1.5 Reports required on an annual basis shall include:

a. A tabulation on an annual basis of the number of station utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associajed man rem exposure according to work and job functions, e.g.,

reactor operations and survei llance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assi gnment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 205 of the individual total dose need not be accounted for . In the aggregate, at least 805 of the total whole body dose received from external sources shall be assi gned to specific major work functions.

b. The complete results of steam generator tube inservice inspec-tions performed during the report period (reference Specification 4.4.5.5.b.)

1 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. 2 This tabulation supplements the requirements of 20.407 of 10 CFR Part 20. ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.6 Routine radiological environmental operating reports coverino the operation of the unit during the previous calendar year shall be 'submitted prior to May of each year. The initial report shall be 1 submitted prior to May of the year following initial criticality. 1 6.9.1.7 The annual radiological environmental operating reports shall include summaries, interpretations, and statistical evaluation 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 surveil-lance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the land use censuses required by Specification 3.13.2. 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 Table 6.9-1 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 missi.ng data shall be submitted as soon as possible in 1 supplementary report. The reports shall also include the following: a summary description of the radiological envivonmental monitoring program including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equip-ment used;a map of all sampling locations keyed to a table giving distances and directions from one reactor; the result of land use census required by the Specification 3. 12.2; and the results of participation in the guality Assurance Program required by Specification 3.12.3. SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.9.1.8 Routine radioactive effluent release reports covering the operating 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. -A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all uni ts at the station; however, for units with separate radwaste systems, the submittal shall specify thd releases of radioactive material from each unit. T OLE fi.9-LrNVIAONMENTALAADIOLOGICAL IIOW)TOAINQ PAOQAALl SUMLIAAY Name of Facility Oocnat No. Local)on of Facility Ao porllnII Petlnd ICnunly, Sl ~ I ~ ) Tyler and Lower Llmll tbsudrer ul Medium or I'erhway Torsi tbrrnl~r ot Alt lndlcasor I.ucerlon wilts lllpb~ sl Annual Ltesn Conlrol Lucgrlons Is~cation) RC)10RThllLE Semi%ad ol Analyses Aeracrlon hie en tlsrrre Mean Ill hkan Il) )Unit ol Measure<nuns) Per for nuut 'LLn) llsnpai)i It)stance anal I)tracnon Ilanpa ls Aenpa OI.CUf)ALlfCES 'Q c 0 ar r, n. ~I a A E sr srl tlomlnel Lower Llrull ot Oesecrlan ILLA)asdellned ln sable nosaslon a. ot Tsbl ~ 4.12 1 ol SpeclllcarlonuLI2.1.1. b Mean and ringo bes~t upon derecsable meeeureurenls only. pression ot delectable msasuranranls al spacllled locarlons ls Initlceled In parenlbeses. Il) d ttos ~ s Tb ~ auarnple dale sre provbled lor llluslr~ siva purposes only. 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 Qstes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-lhter-Cooled Nuclear Power Plants," with data summarized on a quarterly basis following the format of Appendix 8 thereof. The radioactive effluent release reports shall include a summmary of the meteorological conditions concurrent wi th the release of gaseous effluents during each quarter as outlined in Regulatory Guide 1.21, with data summarized on a quarterly basis following the format of Appendix E thereof. The radioactive effluent release reports shall include an assessment of the radiation doses from radioacti ve effluents to individuals due to their activities inside the unrestricted area boundary(Figure 5.1-1) during the report 'period. All assumptions used in making these assessments (e.g., specific acti vity, exposure time and location) shall be included in these reports. The radioactive effluent release reports shall include the following information for all unplanned releases to unrestricted area of radio-active materials in gaseous and liquid effluents: I

a. A description of the event and equipment involved.

b. Cause (s) for the unplanned release.

c. Actions taken to prevent recurrence.

d. Consequences of the unplanned release.

The radioacti ve effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21. In addition, the si te boundary maximum noble gas gamma air and beta air doses shall be evaluated. The meteorological condi tions concurrent with the releases of effluents shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Offsi te Dose Calculation Manual (00CM) . ADMINIS TRATIVE CONTROLS MONTH LY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience shall be submi tted on a monthly basis to the Director, Office of Manage-ment and Program Analysis, U. S . Nuclear Regulatory Commission, hhshington, D. C. 20555, with a copy to the Regional Office of Inspection and Enforce-ment, no later than the 15th of each month following the calendar month covered by the report. In addition, any changes to the Offsite Dose CAl-culational Manual of Specification 6. 15 shall be submitted wi th the Monthly Operating Report wi thin 90 days in which the change (s) was made effecti ve. REPORTA 8 E OCCURRENCES 6.9.1.11 The REPORTABLE OCCURRENCES of Specification 6.9.1.12 and 6.9.1.13 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be com-pleted and reference shall be made to the original report date. PROMPT NOTIFICATION WITH WRITTEN 'FOLLOWUP 6.9.1,12 The types of events listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mai lgram, or facsimile transmission to the Director of the Regional Office, or his designate no later than the first working day following the event, with a written followup report within 14 days. The written followup report shall in elude, as a minimum a completed copy of a license event report form. Information provided on the licensee event report form shall be supple-mented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

a. Failure of the reactor protection system or other systems, subject to limiting safety system settings to initiate the required protective function by the time a monitored para-meter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the requi red protective function.

b. Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the technical specifications.

c. Abnormal degradation discovered in fuel cladding reactor coolant pressure boundary, or primary containment.

ADMINI STRAT IVE CONTROLS Reactivity anomalies involving disagreemnt with the predicted value of reactivity balance under steady state conditions dur-ing power operation greater than or equal to.l/ dk/k; a cal-culatedd. reactivity balance 'ndicating a SHUTDOWN MARGIN less conservative than specified in the technical specification; short-term reactivity increases that, correspond td a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5/ ak/k; or occurrence of any unplanned criticality.

e. Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system(s) used to cope with accidents analyzed in the SAR.

Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional require-ments'of systems required to cope with accidents analyzed in the SAR.

g. Conditions arising from natural or man-made events that, as a direct result of the event require unit shutdwon, operation of safety'systems, or other protective measures requried by technical specifications.

Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bses for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analysis in the safety analysis report or technical specification bases; or discovery during unit life of condi tons not specifically considered in the safety analysis report or technical specifica-tions that require remedial action or corrective measures to prevent the existence or development of an unsafe condition. Occurrence of an unusual or important event that has a significant environmental impact or that has potential enviro-mental impact from unit operation. Conditions where radioactive material contained in liquid or gaseous holdup tanks is in excess of that permitted by the limiting condi-tion for operation established in the technical specifications. TABLE 6.9-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Water Airborne Particulate Fish Milk Vegetables Analysi s (pCi/1) or Gases (pCi/m ) (pCi/Kg,wet) (pCi/1) (pCi/Kg, wet) 3 X104 Mn-54 1 x 10 3X10 Fe-59 4 X 102 1 X104 Co-58 1X10 3X10 Co-60 3X10 1 x 10 Zn-65 3 X 10 2 X 104 Zr-Nb-95 4X10 I -131 0.9 1 X 10' CS-134 30 10 1 x 10 60 xlo Cs-137 50 20 2X10 70 2X10 Ba-La-140 2 X 10 3X10 ADMI N I STRATI YE CONTROLS HIRTY DAY WRITTEN REPORTS .9.1.13 The types of events listed below shall be the subject of critten reports to the Director of the Regional Office within thirty ays of occurence of the event. The written report shall include, as minimum, a completed copy of a licensee event report form. Informa-tion provided on the licensee event report form shall be supplemented, s needed, by additonal narrative material to provide complete explana-tion of the circumstances surrounding the event.

a. Reactor protection system or engineered safety feature in-strument settings which are found to be less conservative than those established by the technical spe ifications but which do not prevent the fulfillment of the functional requirements of affected systems.

b. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown re-quired by a limiting conditon for operation.

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

d. Abnormal degradation of systems other than those specified in 6.9.1.12.c above designed to contain radioactive material resulting from the fission process.

e. An unplanned offiste 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.

f. treasured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 6.9-2 when average over any calendar quarter sampling period. When more than one of the radionuclides in Table 6.9-2 are detected in the sampling medium, this report shall be sub-mitted if:

concentration 1) + concentration 2) = ..'.. > 1.0 limit level 1) limit level 2 When radionuclides other than those in Table 6.9-2 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 Specifica-tions 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 AOMINISTRATIVE CONTROLS 6.9.1.13 cont.

f. result of plant effluents; however, in such an event, the con-dition shall be reported and described in the Annual Radiological Environmental Operating Report.

KECJIII 0 6.9.2 Special reports shall be sumitted to the Oirector of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. Inservice Inspection Program Review, Specification 4.4.10.

b. ECCS Actuation, Specifications 3.5.2 and 3.5.3

c. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.

d. Inoperable Meteorological Monitorina Instrumentation, Specifi-cation 3.3.3.4.

e, liaste Gas Hold-Up System explosive gas monitors, Specification 3.3,3.9,

f. Auxiliarv Building ventilation radiation monitoring,'pecifica-.

tion 3.Z.3.9.

g. Condenser Evaluation System, Specification 3,3.3.9.

DNINISTRATIVE CONTROLS 6.10 RECORD RETENTION In addi'tion to the applicable record retention requirements of Title 10, Code of Feeral Regulations, the following records shall be retained for t least the minimum period indication. 6.10.1 The following records shall be, retained for at least five yeras:

a. Records and logs of unit operation covering time interval at each power level.

b ~ Records and logs of principal maintenance activities, inspec-tions, repair and replacement of principal items of equipment related to nuclear safety.

c. ALL REPORTABLE OCCURRENCES submitted to the Commissions

d. Records of surveillance activities, inspections and cali-brations required by these Technical Specifications.

e. Records of changes made to the procedures required by Specification 6.8.1.

Records of radioactive shipments.

g. Records of sealed source and fission detector leak tests and results.

h. Records of annual physical inventory of all sealed source material on record.

. 10.2 The following records shall be retai ned for the duration of the nit Operating License:

a. Records and drawing changes reflecting unit design modifica-tions made to systems and equipment described in the Final Safety Analysis Report.

b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

c. Records of radiation exposure for all individuals entering radiation control areas.

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

ADMINISTRATIVE CONTROLS E. Records of transient of operational cycles for those unit com-ponents identi'fied in Table 5.7-1.

f. Records of reactor tests and experiments.

g. Records of training and qualification for current members of the unit staff.

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

Records of guality Assurance activities required. by the gA Manual.

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

k. Records of meetings of the PNSRC and the NSDRC 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR.Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

~2 A~22 A 6.12.1 In liew of the "control device" of "alarm signal" required by paragraph 20.203 (c) (2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit*. Any individual or group of in-dividuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

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

" Hea th Physics personnel or personnel escorted by Health Physics per-sonnel in accordance with approved emergency procedures shall be exempt from the R!AP issuance requirement during the performance of their radia-tion protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas. ADMINI STRATI >/E CONTROLS

b. A radiation monitoring device which Continuously integrates the radiation dose rate in the area and alarms when a preset inte-grated dose is received. Entry in to such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowl-edgeable of them.

c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit.

. 12.2 The requi rements of 6. 12. 1, above, also apply to each high radia-ion area in which the intensity of radiation is greater than 1000 mrem/hr. In addition, locked doors shall be provided to prevent, unauthorized entry nto such areas and the keys shall be maintained under the administrative ontrol of the Shift Supervisor on duty and/or the Plant Health Physicist. ADMINISTRATIVE CONTROLS .15 OFFS ITE DOSE CALCULATION MANUAL ODCM UNCTION .15.1 The ODCM shall describe the and parameters to be used 'n the calculation of offsite doses methodologydue to radioactive gaseous and liquid effluents and in the calculation of gaseous arid liquid effluent monitoring 'nstrumentation alarm/trip setpoints consistent with the applicable LCO's ontained in these Technical Specifications. . 15.2 Any changes to the ODCM shall .be made by either of the following ethods: A. Licensee initiated changes:

1. Shall be submitted to the Commission by inclusion in the Monthly Operating Report pursuant to Specification 6.9.1.10 within 90 days of the date the change(s) was made effective and shall contain..

a. sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental i nformation. 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 both the PNSRC and NSDRC.

2. Shall become effective upon a date specified and agreed to by both the PNSRC and NSDRC following their review and acceptance of the change (s),

B. Commission initiated changes:

1. Shall be determined by the PNSRC to be applicable to the facility after consideration of facility design.

ADMINISTRATIVE CONTROLS

2. The licensee shall provide the Commission with written noti-fication of their determination of applicability includino any'necessary revisions to reflect faci'lity design.

3. Shall be reviewed by the (NSDRC) at its next regularly scheduled meeting.

MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (liquid, Gaseous and Solid FUNCTION 6.16.1 The radioactive waste treatment systems (liquid, gaseous and solid) <<are those systems described in the facility Final Safety Analysis Report ..or Hazards Summary Report, and amendments thereto, which are used to ',maintain that control over radioactive materials in gaseous and liquid >effluents and in solid waste packaged for offsite shipment required to 'meet the LCO's set forth in Specifications 3.11.1.1, 3.11.1.3, 3.11.1.4, 3.11.1.4, 3.11.2.1, 3.11.2.2, 3.11.2.3, 3.11.2.4, 3.11.2.5, 3.11.2.7, and 3.11.3.1. 6.16.2 Major changes to the radioactive waste systems (liquid, gaseous and solid) shall be made by either of the following methods, For the purpose of this specification 'major changes's defined in Specifica-tion 6.16.3 below. A. Licensee initiated changes: The Commission shall be informed of all changes by the in-clusion of a suitable discussion of each change in the Semiannual Radioactive Effluent Release Report for the period in which the changes were made. The discussion of each change shall contain:

a. a summary of the evaluation that led to the determina-tion 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 additi:ooal or supplemental information.

c. a detailed description of the equipment, components and processes involved and the interfaces wi th other plant systems;

ADi~ISTRATIVE CC5TROLS an evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste from those previously predicted in the licensee 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 from those previously estimated in ".he licensee 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 in which the changes were 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 both the (PNSRC) and the (NSDRC).

2... The change shall become effective 'upon review and acceptance by both the (PNSRC) and (NSDRC). B. Commission initiated changes:

1. The applicability of the change to the facility shall be determined by the (PNSRC) after consideration of the facility design.

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

3. The change shall be reviewed by the (NSDRC) at its next regularly scheduled meetings

4. The change shall become effective on a date specified by the Commission.

6.16.3 Background and definition of what constitutes 'major changes'o radioactive waste systems ( liquid, gaseous and solid) ~ ADt1INISTRAT I VE CONTROLS A. 'ackground 10 CFR Part 50, Section 50.34a(a) requires that each application to construct a nuclear power reactor provide a description of the equipment installed to maintain control over radioactive material in gaseous and liquid effluents produced during normal reactor operations in-cluding operational occurrences.

2. 10 CFR Part 50, Section 50.34a (b) (2) requires that each application to construct a nuclear power reactor provide an estimate of the quantity of radionuclides expected to be released annually to unrestricted areas in liquid and gaseous effluents produced during normal reactor operation.

3 ~ 10 CFR Part 50, Section 50.34 a(3) requires that each application to construct a nuclear power reactor provide a description of the provisions for packaging, storage and shipment offsite of solid waste containing radio-active materials resulting from treatment of gaseous and liquid effluents and from other sources. 10 CFR Part 50, Section 50.34a (c) requires that each application to operate a nuclear power reactor shall in-clude (1) a description of the equipment and procedures for the control of gaseous and liquid effluents and for the maintenance and use of equipment installed in radioactive waste systems and (2) a revised estimate of the information required in (b)(2) if the expected releases and exposures differ significantly from the estimate submitted in the application for a construction permit.

5. The Regulatory staff's Safety Evaluation Report and amend-ments thereto issued prior to the issuance of an operating license contains a description of the radioactive waste systems installed in the nuclear power reactor and a de-tailed evaluation (including estimated releases of radio-active materials in liquid and gaseous waste and quantities of solid waste produced from normal operation, estimated annual maximum exposures to an individual in the unres-tricted area and estimated exposures to the general popula-tion) which shows the capability of these systems to meet the appropriate regulations.

ADMINISTRATIVE CONTROLS

6. The Regulatory staff's Final Environmental Statement issued prior to the issuance of an operating license con-tains a detailed evaluation as to the expected environ-mental impact from the estimated releases of radioactive material in liquid and gaseous effluents, Definition "Major Changes" to radioactive waste systems (liquid, gasesous and solid) shall include the following:

1. Major changes in process equipment, components, structures and effluent monitoring instrumentation from those described in the Final Safety Analysis Report (FSAR) or the Hazards Summary Report and evaluated in the staff's Safety Evalua-tion Report (SER) ( e.g., deletion of evaporators and installation of demineralizers; use of fluidized bed calciner/incineration in place of cement solidifcation systems);

• 2. Major changes in the design of the radwaste treatment system (liquid, gaseous and solid) that could significantly alter the characteristics and/or quantities of effluents released
,-or volumes of solid waste stored or shipped offsite from those previously considered in the FSAR and SER (e.g., use of asphalt system in place of cement);-

3. Changes in system design which may invalidate the accident analysis as described in the SER (e,g. changes in tank capacity that would alter the curies released); and Changes in system design that could potentially result in a significant increase in occupational exposure of operating personnel (e.g., use of skid mounted equipment, use of mobile processing equipment).