Text

Proposed Tech Specs Re Radiological Effluents
	ML20235F768
Person / Time
Site:	Rancho Seco
Issue date:	06/30/1987
From:	; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	;
Shared Package
ML20235F757	List: ML20235F754; ML20235F768;
References
	NUDOCS 8707130538
	Download: ML20235F768 (335)
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FACILITY CHANGE SAFEIT ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 2

1. Existing Specification:

1.13 PROCESS CONTROL PROGRAM l i

A PROCESS CONTROL PROGRAM (PCP) shall be the manual I detailing the program of sampling, analysis, and evaluation l within which SOLIDIFICATION of radioactive wastes from )

liquid system is assured.

1.14 SOLIDIFICATION Solidification shall be the conversion of liquid radioactive wastes to an immobilized free-standing solid.

New Specification:

1.13 PROCESS CONTROL PROGRAM PROCESS CONTROL PROGRAM (PCP) - The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured.

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

Discussion:

The changes here are administrative and add clarification to the existing definitions. There is no technical variation in meaning for the Process Control Program or Solidification.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 3

2. Existing Specification:

1.15 0FFSITE DOSE CALCULATION MANUAL (ODCM)

An OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be a manual containing the methodology and parameters to be used in the-calculation of offsite dose due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints and specific details of the environmental radiological monitoring program.

New Specification:

1 1.15 0FFSITE DOSE CALCULATION MANUAL (ODCM)

The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be a manual containing the description of the methodology, algorithms and parameters to be used in the calculation of offsite doses resulting from the release of radioactive material in gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints.

Discussion:

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The details of the radiological environmental monitoring program were originally included in the ODCM. Under these tech spec changes the monitoring program will be a separate document from the ODCM and delineated and t.itled the Radiologieni Environmental Monitoring Program (REMP) manual. There is no District commitment to the NRC to separate the ODCM/REMP definitions, but it is in the District's best interest to do so because it improves the overall understanding of the District's effluent release program.

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'FACILITYCHANGESkFETYANALYSIS LOG No. 921

l. ~ PROPOSED AMENDMENT NO. 155 PAGE 3a l

2a. Existing Specification:

1.17 SITE BOUNDARY i

The boundary of the SMUD property.

New Specification: !

1.17 SITE BOUNDARY Site Boundaries are defined by Figure 5.1-1 through 5.1-4.

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Discussion:

Figures 5.1-1 through 5.1-4 show the details of the SMUD owned property which form the site boundary.

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. FACILITY CHANGE SAFErf ANALYSIS LOG No. 921-PROPOSED AMENDMENT No. 155 PAGE 4

- 3. . Existing Specification:

1.18' DOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844,

" Calculation of Distance Factors for Power and Test Reactor Sites."

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New Specification: !

l 1.18 DOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of l I-131 (microcurie / gram) which alone would produce the same l thyroid dose via the inhalation pathway as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP Publication 30, " Limits for Intakes of Radionuclides by Workers," 1979.

{ l Discussion: l The new specification provides clarification of the thyroid dose equivalent pathway of exposure through inhalation. There had been some previous difficulties as to what constitutes " dose equivalent I-131." The changes clarify the definition with the use of the latest scientific l' information available documented in ICRP Publication 30, " Limits for Intake of Radionuclides by Workers," 1979.

1 FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 l PROPOSED AMENDMENT NO. 155 PAGE 5

4. Existing Specification:

l N/A New Specification: )

l 1.21 MAXIMUM HYPOTHETICAL INDIVIDUAL The LAXIMUM HYPOTHETICAL INDIVIDUAL is characterized as

" mas.imum" with regard to food consumption, occupancy, or j other usage or exposure pathway parameters in the vicinity l of Rancho Seco that would represent an individual or composite of individuals with habits greater than usually expected for the average of the population in general. No single individual would be expected to be exposed to all the potential pathways at the " maximum" value.

The MAXIMUM HYPOTHETICAL INDIVIDUAL is a hypothetical receptor of radiological exposure (mrem) resulting from the discharge of radioactive effluent (curies). The methodology to convert curies into mrem is described in the ODCM. \The purpose of the ODCM calculation is to compare the resultant effluent exposure or dose with the numerical guides for design objectives in 10 CFR 50, Appendix I.

The MAXIMUM HYPOTHETICAL INDIVIDUAL concept is consistent with its use in the US NRC Regulatory Guide 1.109 and 10 CFR 50, Appendix I.

The MAXIMUM HYPOTHETICAL INDIVIDUAL concept is NOT used to demonstrate compliance with 10 CFR 20.106.

1.22 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGPJW (REMP) MANUAL The RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

MANUAL shall be a manual containing the description of the Rancho Seco radiological environmental monitoring program.

The REMP manual shall contain a description of the environmental samples to be collected, the sample locations, sampling frequencies, and sample analysis criteria.

1.23 LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM The LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM is the system designed and installed to reduce radioactive materials in liquid effluents by collecting liquid effluent and providing processing for the purpose of reducing the total radioactivity prior to release to the environment.

1 FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 6 i

4. New Specification: (Cont.) l l

1.24 VENTILATION EXHAUST TREATMENT SYSTEM l l

The VENTILATION EXHAUST TREATMENT SYSTEMS are systems designed and installed to reduced gaseous radioiodine or l radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal

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l absorbers and/or HEPA filters for the purpose of removing l iodines or particulate from the gaseous exhaust stream I prior to the release to the environment (such a system is j not considered to have any effect on noble gas effluents). I Engineered Safety Feature (ESF) atmospheric cleanup systems 'j are not considered to be VENTILATION EXHAUST TREATMENT SYSTEMS components. ,

1. 25 PURGE - PURGING j i

PURGE or PURGING is the controlled process of discharging l air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. j

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

1.27 RADIOACTIVE EFFLUENT l

Effluent shall be designated as RADI0 ACTIVE EFFLUENT when !

the radiochemical analysis of an appropriate sample of the effluent results in the detection of radioactive material above the Lower Limits of Detection as defined in the OFFSITE DOSE CALCULATION MANUAL.

l Discussion: i The USNRC Regulatory Guide 1.109 is the acceptable methodology for calculating dose resulting from the discharge of radioactive materials in gaseous and liquid effluents for the purposes of comparison with the numerical guides for design objectives of 10CFR50, Appendix I. The j definitions of Maximum Hypothetical Individual (MHI) is necessary to {

differentiate methodologies for measuring compliance with 10CFR50 Appendix 4

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4 FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 7

4. Discussion (Cont.)

I. The MHI definition is pursuant to 10CFR50, Appendix I dose methodology and is more conservative' than a real dose to a Member of the Public. The definition for the Radiological Environmental Monitoring Program (REMP) manual is provided as a separate document from the ODCM. The remaining definitions are added per District commitment to the NRC.

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1 FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 8

5. Existing Specification:

'3.15 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION _

The radioactive liquid effluent minitor'ing instrumentation channels shown in Table 3.15-1 sha'l be OPERABLE with.their l alarm / trip setpoints set to ensure thet the limits of Specification 3.17 are not exceeded.

Applicability .During radioactive releases via the pathways identified in Table 3.15-1.

Action

a. With a radioactive liquid effluent monitoring ;

instrumentation channel alarm / trip setpoint less conservative than a valve which will ensure that the limits of Specification 3.17 are met, without delay suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable or change the setpoint so it is aheeptably conservative. ;

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b. With less than the minimum number of radioactive j liquid effluent monitoring instrumentation channels j OPERABLE, take the ACTION shown in Table 3.15-1. j Bases During normal operations, all radioactive contaminated water from primary system leaks and drains is processed in a liquid radwaste system and recycled into the Reactor Coolant Makeup System or otherwise reused in the controlled arcas of the plant. Only secondary system water is normally released'from the plant. The secondary system water, if contaminated, would be released through the Regenerant Hold-Up Tanks.

During periods of primary to secondary leakage, or when the sumps are contaminated, administrative controls require the turbine building sumps liquid effluent to be diverted to the Regenerant Hold-Up Tanks.

Under normal conditione, the once through steam generators have no blow down. If a blow down is required during periods of primary to secondary leakage, all water will be retained and processed in the radwaste system or diverted to the Regenerant Hold-Up Tanks. I Upon indication of radioactivity in the secondary system, radioactive liquid effluent instrumentation is required to monitor and control, as applicable, the releases of radioactive materials in liquid

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 9

5. Existing Specification: (Cont.)

effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

New Specification:

3.15 RADIOACTIVE LIQUID EFFLUENT MONITORING The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.15-1 shall be OPERABLE with their alarm / trip setroints set to ensure that the limits of specification 3.17.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the methodology contained in the OFFSITE DOSE CALCULATION MANUAL (0DCM).

Applicability During the release of radioactive effluents via the pathways identified in Table 3.15-1.

Action

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less ,

conservative than a value which will ensure that the limits of Specification 3.17.1 are met, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.15-1.

Exert best efforts to return the instrument to OPERABLE status within 30 days and, if unsuccessful, l explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.2.3 why the inoperability was not corrected in a timely manner.

Bases During normal operations, radioactive contaminated water from primary system leaks and drains is processed in a liquid radwaste system and

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1 FACILITY CHANGE SAFErf ANALY3IS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 10

5. New Specification:"(Cont.) i

. recycled into the Reactor Coolant Makeup System or otherwise reused in the controlled areas of the plant. Secondary system water is-normally released from.the plant. J t

The secondary system water, if it contains-radioactive' material is:

released through the~'A' and 'B' Regenerant Hold-Up Tanks (RHUTs).

.During periods of primary to secondary leakage, or when the sumps are contaminated, administrative controls require the turbine building .

sumps liquid effluent to be diverted to the 'A' and 'B' Regenerant Hold-Up Tanks.

Demineralized reactor. coolant can be transferred from the Demineralized Reactor Coolant Storage Tank (DRCST) to the 'A' and 'B'-

Regenerant Hold-Up Tanks for sampling, processing and eventual-discharge offsite as' required by operational constraints.

Under normal conditions, the once through steam generators have no blow down. If a blow down is required during periods of primary to secondary leakage, all water will be retained and processed in the radwaste system or diverted to the 'A' and 'B' Regenerant-Hold-Up Tanks. \

~ Radioactive liquid effluent instrumentation is provided to monitor and. control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the methodology contained'in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.106. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60,.63, and 64 of Appendix A to 10 CFR Part 50.

Discussion The changes incorporate the provisions documented in the Standard. )

Radiological Effluent Technical Specifications (RETS). Clarification is i made that the primary water system of the plant may be released from the site via the Regenerant Hold-Up Tanks (RHUTs) 'A' and 'B'. Additional clarification to the Tech Specs state that the ODCM contains the-methodology to calculate the setpoints of effluent monitoring instrumentation, and not the implementing procedures.

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6. Existing Specification: Page 11 Table 3.15-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum

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Number of Channels Instrument . Operable -Action

1. Gross Radioactivity Monitors Providing Auto- -

matic Termination of 1 Release -

a. Regenerant Hold-Up 1 With the monitor inoperable, Tank Discharge Line effluent releases may be resumed provided that prior to initiating a Monitor '

release:

1. At Teast two independent samples !

are analyzed in accordance with g

Specification 3.17. .

2. A second member of the- facility

) technical or operational staff will independently verify the

. release rate calculations and discharge valving.

3. Exert best efforts to return the instrument to OPERABLE 1 status within 30 days and, if unsuccessful, explain in the l next Semiannual Radioactive Effluent Release Report why the I inoperability was not corrected in a. timely manner.

2. FTow Rate Measurement Devices

a. Regenerant Hold-Up Tank 1 With the fTow rate measurement

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Discharge Line Monitor device inoperable, effluent releases via this pathway may continue provided the flow rate is '

estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump performance curves generated in j situ may be used to estimate flow.

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6. Existinn Specification: (Cont.)

Page 12 -- ;

Table 3.15-1 (Continued)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ,

Minimus Number of Channels Instrument Operabid Action

2. Flow Rate Measurement Devices (Continued)

b. Waste Water Flow 1 With the flow rate measurement device inoperable, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

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6. New Specification: Page 13

. Table 3.15-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable Action

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Retention Basin 1 With the monitor inoperable, Effluent Discharge effluent releases may be resumed Monitor provided that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Specification 4.21.1.

2. At least two technically

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qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of .

radioactive effluents via this pathway.

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6. New Specification: (Cont.) Page 14

.t Table 3.15-1 (Continued) f I

RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Number i

of Channels Instrument Operable Action j

2. Flow Measurement Devices

a. Regenerant Hold-Up 1 With the flow measurement device Tank Discharge Line inoperable, releases to the retention '

Total Flow basins may continue provided the total flow can be detemined by a tank level . device or pump perfomance curves.

b. Waste Water Flow Rate 1 With the flow rate measurement device inoperable, effluent releases via this pathway may continue provided the flow rate is estimated i

\ at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during retention basin releases.

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FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 15

6. New Specification: (Cont.)

Discussion The addition of the Retention Basin discharge monitor, which will serve as the District's new effluent control point, will provide more effective control of effluent release. The RHUT total flow monitor is edded to assess total curies of radioactive material released offsite for offsite dose calculation. The addition of the RHUT Discharge Line Mon 3. tor specification is provided for measurement of the RHUT volume released to-the retention basin and the determination of total offsite dose.

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FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 16

7. Existing Specification:

3.16 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.16-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.18 are not exceeded.

Applicability During release via the pathways identified in Table l 3.16-1. j Action .I

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of Specification 3.18 are met, immediately suspend the release or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.16-1.

Bases 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 instruments shall be calculated in accordance with the methods in the ODCM to ensure that the alarm / trip ,

will occur prior to exceeding the limits of 10 CFR Part 20. The ?

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

The Waste Gas Header Monitor monitors the Waste Gas Holdup System noble gas releases and will provide automatic termination of the release. However, it is located on the system header and monitors the noble gas prior to dilution in the Auxiliary Building ventilation system and passing through HEPA and charcoal filters. The Auxiliary Building Stack alarms and terminates the release automatically if it exceeds the limits. Therefore, as the Auxiliary Building Stack is the effluent release point and will perform the necessary Waste Gas Holdup System release termination, it is listed as the Technical Specification instrument.

FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 ,PAGE 17

7. Existing Specification: (Cont.)

The air ejector exhaust and gland seal exhaust also have individual noble gas monitors. These systems exhaust into the Auxiliary Building ventilation system. Therefore, as the Auxiliary Building Stack is the effluent release point and will alarm if either of these systems release environmentally significant gases, it is used as the Technical Specification instrument.

New Specification:

3.16 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.16-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.18.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the methodology contained in the ODCM.

Applicability During release via the pathways identified in Table 3'.16-1.

Action

a. With a radioactive gaseous effluent aonitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of Specification 3.18.1 are met, immediately suspend the release of radioactive gaseous effluent monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.16-1.

Exert best efforts to return the instrument to OPERABLE status within 30 days and, if unsuccessful, erplain in the next Sem1 annual Radioactive Effluent Report pursuant to Specification 6.9.2.3 why the inoperability was not corrected in a timely manner.

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The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents .during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments

FACILITY CHANGE SAFEiT ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 18

7. New Specification: (Cont.)

shall be calculated in accordance with the methodology contained in-the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.106. The OPERABILITY and use of-this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The Auxiliary Building Stack is the effluent release point for the Waste Gas System and the Auxiliary Building Stack Noble Gas Activity -

Monitor will perform the necessary Waste Gas System. release-termination. The monitor alarms and terminates a Waste Gas Decay Tank release automatically if the activity exceeds the setpoint limits.

~The condenser air ejector exhaust has an individual noble gas monitor. This system exhausts'into the Auxiliary Building ventilation system. 'Therefore, the Auxiliary Building Stack is the effluent release point and will alarm upon release of environmentally significant radioactive gases.

Fuel Storage Building exhaust is directed to the Auxiliary Building stack where the' exhaust will be filtered and monitored for any activity prior to being released to the atmosphere.

Discussion:

The changes represents conformance with the. Standard RETS. Additional clarification is made that the methodology to determine the setpoints for the radioactive gaseous effluent instrumentation is only contained in the ODCM.

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8. Existing Specification: Page 19 .

Table 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number

- . of Channels '

Instrument Operable Action

1. Reactor Building Purge .

Vent

a. Noble Gas Activity 1 With the monitor channel alarm /

Monitor providing trip setpoint less conservative alam and automatic than required by Specification temination of 3.16, imediately st. spend the release release or declare the channel inoperable.

With the monitor inoperable, ,

effluent releases via this . ' ;

pathway may continue provided )

grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these g

samples are analyzed in accordance with Table 4.22-1

- within 2.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ,

b. Iodine Sampier- -1 With the co1Tection device

. inoperable, effluent releases via this pathway may continue provided continuous samples. are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. Particulate Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with' Table 4.22-L within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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8. Existing Specification: (Cont.) Page 20 Table 3.16-1 (Continued) .

,RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum

- Number ,

of Channels Instnment Operable Action

1. Reactor Building Purge Vent (continued) ,

d. System Effluent Flow 1 With the flow rate device Rate Device inoperable, effluent releases may continue providea the flow rate used is the maximum design flow rate.

e. Sampler Flow Rate L With the flow ra'te device Measurement Device inoperable, effluent releases.

via. this pathway may continue provided the flow rate is estimated and recorded at least g

once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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% 8. Existing Specification: (Cont.) Page 21

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' Table 3.16-1 (continued)

RADIOACTIVE GASES EFR.UENT MONITORING INSTRUMENTATION Ministas Number

. of Channels Instrument Operable Action

2. Auxiliary Building Stack t a. .t Noble Gas Activity . 1 With the monitor channel alarm / ,

Monitor providing trip setpoint less conservative than required by Specification

.h 3 alans and automatic 3.16, imediately suspend the j ' tennination of release or declare the channel release c'

inoperable.

i With the monitor inoperable, effluent releases via this pathway may continue provided grah samples are taken at least once per 12. hours and these 3

\ samples are analyzed in accordance with Table 4.22-1 withf tr 24. hours.

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b. Iodine Sampler I Witit the collection device inoperable, effluent releases via this pathway may continue

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t' prfpided continuous samples are takert and these 5:imples are ,

analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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' c. Particulate Sampler 1 With the collection device inoperable, effluent releases 4, '

via this pathway may continue

' pnwided continuous samples are taken and these samples are analyzed in accordance witir

TabTe 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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8.'Existini Specification: (Cont.)

Page'22

. Table . 3.16-1. (Continued) ,

i RADI0 ACTIVE GASES CFFl.UENT MONITORING INSTRUMENTATION

)

Minimum l Number of Channels Instrument Operable Action .

2.- Auxiliary Building

. Stack (continued)

d. . . System Effluent Flow 1 1With the flow rate device Rate Device inoperable, effluent releases' via this pathway may' continue provided the flow rate used is the maximum design ' flow rate.

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e. Samp ar Flow Rate 1 With the flow rate device Measuring Device inoperable, effluent releases l via this pathway may continue provided the flow rate.is

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estimated and reccrded at 1 east once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

f. Wasta Gas. Holdup 1. With the monitor channel alann/

Systes (Auxilfary trip sotpoint less conservative-Building Stack than required by Specification Monitor) - ; 3.16, famediately suspend the j release or declare the channel

- inoperable.

With the monitor inoperable, the contents of the tank (s) may be released to the environment provided that prior to initiating the release:

a. At least two independent l samples of.the tank's contents are analyzed, arid

- b. At least two technically qualiffed members of the Facility Staff independently verify the release rate

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calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.

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8. Existina Specification: (Cont.) Page 23

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l Table 3.16-1 (continued) -

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RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum j Number of Channeis '

Instrument Operable ,

Action

3. Radwaste Service -

Aret Vent * .

a. Noble Gas Activity 1 With the monitor channel alarm /

Monitor trip setpoint less conservative than required by Specification 3.16, immediately suspend the release or declare the channel inoperable.

.With the monitor inoperable.

. effTuent releases via this pathway may continue provided grab samples are takert at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed. in

\ accordance withr Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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b. Todine Sampier 1. With the co1Tection device

- inoperable, effTuent reTeases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Tird Radwaste Service Ares Vent Monitoring Systen is not yet functional .

Tfsis specification for this system will become effective where it is

- dn: Tared OPERA 8t.E. .

e e

e em .

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8. Existing Specification: (Cont.') . Pa;;e 24 Table 3.16-1 (continued) :

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument _ Operable Action 1

3. Radwaste Service ,

Area Vent * (continued)

c. Particulate Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided centinuous samples are taken and these samples are analyzed in accordance with Table 4.22-L within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. System Effluent Flow L With the flow rate device '

Rate Device inoperable, effTuent releases .

may continue provided the flow rate used is the maximum design '

fTow rate. ,

e. Sampler Flow Rab L With the flow rate device Measurement Device inoperable, effluent releases

' f; via. this pathway may continue

provided the fTow rate is

- estimated and recorded at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The Radwaste Service Area Vent Monitoring System is not yet functional.

This specification for this system will become effective when it is declared OPERABLE.

d er 9

8. New Specification: Page 25 l

. Table 3.16-1 ,

RADIOACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable Action

1. Reactor Building Purge Vent

a. Noble Gas Activity 1 With the monitor channel alarm /

Monitor providing trip setpoint less conservative alarm and automatic than required by Specification termination of 3.18.1, immediately suspend the release release or declare the channel

. t . inoperable.

With the monitor inoperable, effluent releases via this pathway may continue provided grab samples g

are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. Iodine Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided .

contiriuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. Particulate Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

r 1 l

8. New Specification: (Cont.)^ Page 26 j l

I

. Table 3.16-1 (Continued) f RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum.

Number of Channels . 1 Instrument -

Operable Action

1. -

Reactor Building) Purge Vent (continued

d. System Effluent Flow 1 With the flow rate device Rate Device inoperable, effluent releases may continue provided the flow rate used is the maximum design flow rate.

e. Sampler Flow Rate 1. With less than the minimum number of Measurement Devices channels operable, effluent relea'ses via this pathway may continue provided the flow rate is estimated and recorded at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1 -

8. New Specification: (Cont.) .Page 27

.f Table 3.16-l' (Continued) I RADIOACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION

- Minimum Number of Channels Instrument Operable ,_ Action

2. Auxiliary Building Stack

a. Noble Gas Activity 1 With the monitor inoperable, Monitor providing effluent releases via this pathway alarm may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l

b. Iodine Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and

\

these samples are analyzed in accordance with Table 4.22-1 within 0 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. Particulate Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within-24 hours.

_ __.A__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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8. New Specification: (Cont.). Page 28 Table 3.16-1, (Continued)

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels- i Instrument Operable Action

2. Auxiliary Building Stack (continued)

d. System Effluent Flow 1 With the flow rate device Rate Device inoperable, effluent releases via this pathway may continue provided the flow rate used is the maximum design flow rate.

e. Sampler , Flow Rate 1 With the flow rate device inoperable, Measuring Devices effluent releases via this pathway may continue provided the flow rate is estimated and recorded at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

f. Waste Gas \ 1 With the monitor channel alam/

System (Auxiliary trip setpoint less conservative Building Stack than required by Specification 5 Monitor) 3.18.1, f amediately suspend the release or declare the channel inoperable.

the With the monitor contents of the tank inop(erable s) may be released to the environment provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.

4

'8. New Specification: (Cont.) Page 29 Table 3.16-1 (Continued)

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable Action

3. Auxiliary Building Grade Level Vent

a. Noble Gas Activity 1 With the monitor channel alarm /

Monitor trip setpoint less conservative than required by Specification 3.18.1, immediately suspend the release or declare the channel inoperable.

With the monitor inoperable, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed in

\ accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. Iodine Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

t

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8. New Specification: (Cont.) Page 30 Table 3.16-1 (continued)

RADIOACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instruraent Operable _ Actio'n

~

3. Auxiliary ' Building Grade Level Vent (continued) l

c. Particulate Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. System Effluent Flow 1 With the flow rate device Rate Device inoperable, effluent releases may continue provided the flow rate used is the maximum design

\

flow rate.

e. Sampler Flow Rate 1 With the flow rate device 5 Measurement Device inoperable, effluent releases via this pathway may continue provided the flow rate is estimated and recorded at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 31

8. New Specification: (Cont.)

Discussion:

Setpoints for the noble gas activity monitor for the Reactor Building l Purge Vent, the Auxiliary Building Grade Level Vent and the Auxiliary

~

Building Stack are based on compliance with 10CFR20 requirements as specified in Technical Specification 3.18.1. Additional changes are based on compliance with Standard RETS. j 4

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 32 l

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9. Existing Specification: f 4

3.5.7 LIQUID EFFLUENTS i l

3.17.1 CONCENTRATION The concentration of radioactive material released at any time beyond the site boundary shall be limited to the concentrations specified in 10 CFR 20, Appendix B, 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 2X10-4 uCi/ml.

Applicability At all times.

Action With the concentration of radioactive material released from the site to unrestricted areas exceeding Specification 3.17.1, restore concentration within the specification limits as soon as practicable.

Bases \

This Specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to areas beyond the site boundary will be less than the concentration levels specified in 10CFR Part 20, Appendix B, Table II. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not j result in exposures within: (1) the Section II. A Design Objectives '

of Appendix I, 10CFR Part 50, to an individual, and (2) the limits of 10CFR Part 20.106 (e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotopes and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in international Commission on Radiological Protection (ICRP) Publication 2.

New Specification:

3.17 LIQUID EFFLUENTS 3.17.1 CONCENTRATION The concentration of radioactive material released in liquid effluents at any time beyond the Site Boundary For i Liquid Effluents (see Figure 5.1-4.) shall be limited to the i concentrations specified in 10CFR Part '20, Appendix B, Table II, Column 2 for r. radionuclides other than dissolved i

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 33

9. New Specification (Cont.)

or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2X10-4 uCi/ml total activity.

Applicability At all times.

Action With the concentration of' radioactive material released from the site exceeding Specification 3.17.1, immediately restore concentration within the specification limits and report the event in the next Semiannual Radioactive j Effluent Release Report pursuant to Specification 6.9.2.3.

Bases This Specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the ,

site to areas beyond the Site Boundary For Liquid Effluent (see l Figure 5.1-4) vill be less than the concentration levels specified in 10CFR Part 20, Appendix B, Table II, Column 2. This limitation ;

provides additional assurance that the levels of radioactive

}

materials in bodies of water outside the site will result in exposures within the limits of 10CFR Part 20.106 to MEMBER (S) 0F THE ,

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

Publication 2.

Discussion:

The changes here relate to compliance with the Maximum Permissable Concentration (MPC) of 10CFR20, Appendix B, Table II, Column 2. Addition of Figure 5.1-4 and the action to report in Semiannual Radioactive Effluent Release Report are included for compatibility with the Standard RETS. The existing specification was based on the Standard Radiological Effluent Technical Specifications (RETS) which assumes that for a

1 FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 i PROPOSED AMENDMENT NO. 155 PAGE 34

9. Discussion: (Cont.)

" standard PWR" compliance with the MPC limit on an hour by hour bases will !

also result in the plant operation being ALARA in terms of the numerical l guides for the design objectives of 10CFR50 Appendix I. This is an H incorrect assumption for the site specific environmental setting of Rancho ]

Seco, therefore the Appendix I statement in the LCO has been deleted. j i

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 35

10. Existing Specification:

3.17.2 DOSE The dose or dose commitment to a member ~of the public from radioactive materials in liquid effluents released beyond the site boundary shall be limited:

a. During any calendar quarter 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 or Jose commitment from the release of radioactive material in liquid effluents exceeding any of the above limits, prepare and submit t'o the Commission within 30 days a Special Report.

This Report will identify the cause(s) for exceeding the limit and define the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

Bases This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as reasonably achievable." The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to substantially underestimated. The equations specified in the ODCM for calculation the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Efflueats for

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FACILITY CHANGE' SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 36

]

1 1

10. M rcing Specification: (Cont.)

the Purpose of Evaluating compliance with 10CFR 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.

New Specification:

3.17.2 DOSE The dose or dose commitment to a MAXIMUM HYPOTHETICAL '

INDIVIDUAL from radiological materials in liquid effluents released beyond the Site Boundary For Liquid Effluents (see Figure 5.1-4) shall be limited to:

a. Less than or equal to 1.5 mrem to the total body and to less than or equal to 5.0 mrem to any organ during any calendar quarter; and,

b. less than or equal to 3 mrem to the total body and to Idss than or equal to 10 mrem to any organ during any

calendar year. '

U Applicability At all times.

Action

a. With the calculated dose or dose commitment from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report pursuant to Specification 6.9.5. This Report will identify the cause(s) for exceeding the limit (s) and define the corrective actions to be taken to reduce the releases of radioactive material in liquid effluents and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

Bases l

This specification is provided to implement the requirements of )

Sections II.A, III.A and IV.A of Appendix I, 10CFR Part 50. The l Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the 1

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__ _ - - - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _J

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FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMLNT NO. 155 PAGE 37 I

10. New Specification (Cont.)

required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to. assure that the releases of radioactive material in liquid effluents will be kept "as low as reasonably achievable." The dose calculation methodology in the ODCM implement the >. requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Release of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR 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.

There is reasonable assurance that the operation of the facility will not result in radionuclides concentrations in finished drinking water that are in excehs of the requirements of 40 CFR 141.

Discussion:

The changes represent adoption to Standard Technical Specifications NUREG-0472 Rev. 2, and NUREG-0452, Rev. 5 (draft). NUREG-0472 and NUREG-0452 incorporate provisions which include verifying that measurable ,

concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of environment erposure pathways.

The current Rancho Seco Tech Specs (RSTS) incorrectly references 10CFR50 Appendix I to a real individual (member of the public) where the correct reference is to a Maximum Hypothetical Individual. Figure 5.1-4 is added to reflect current site boundary for liquid effluents.

All other wording changes bring the current RSTS into standardization with Standard RETS which is a commitment to the NRC and is in the District's best interest.

- - - ~ - - - - - - ~ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

FACILITY CHANGE SAFETY AMEYSIS LOG No. 921 PROPOSED AMENDMDIT NO.155 PAGE 38

11. Exicting Specification:

3.17.3 LIQUID HOLDUP TANKS The quantity of radioactive material contained in each of the following tanks shall be limited to less than or equal to 10 Curies, excluding tritium and dissolved or entrained noble gases:

a. Regenerant Holdup Tanks

b. Outside Temporary Tanks Applicability At all times Action With the quantity of radioactive material in any of the listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hou,rs reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report.

Bases Restricting the quantity of radioactive material contained in the specified outdoor tanks provides assurance that in the event of an uncontrolled release of the contents, the concentration at the nearest surface water supply in an unrestricted area would be less than the limits of 10 CFR 20, Appendix B, Table II, Column 2. There are two Regenerant Holdup Tanks. The limit applies to each tank individually.

New Specification:

3.17.3 LIQUID HOLDUP TANKS The quantity of radioactive material contained in each of the following tanks shall be limited to less than or equal to 10 Curies, excluding tritium and dissolved or entrained noble gases:

a. "A" and "B" Regenerant Holdup Tanks

b. Borated Water Storage Tank

c. Demineralized Reactor Coolant Storage Tank

)

l r FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 39

11. New Specification: (Cont.)

d. Miscellaneous Water Holdup Tank

e. Outside Temporary Tanks !

l Applicability At all times Action With the quantity of radioactive material in any of the i listed tanks exceeding the above limits, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report.

Bases I i

The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes,\ or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains g'

connected to the Liquid Radwaste Treatment System or the .

LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM. j Restricting the quantity of radioactive material contained in the specified outdoor tanks provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentration at the nearest potable water supply and the nearest surface water supply ';

in an unrestricted area would be less than the limits of 10 CFR 20, Appendir B, Table II, Column 2. The limit applies to each tank individually.

Discussion The changes identify all the tank outfalls that will contribute to the Rancho Seco liquid effluent. Assurance is made,that the resulting radionuclides concentration in each tank will be less than the limits of 10 CFR 20, Appendix B, Table II, Column 2 so that in the event of an uncontrolled release, the resulting radionuclides concentration in the nearest potable and surface water supply in an unrestricted area will also be less than the aforementioned limits.

' /*.

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. FACILITY CHANGE SAFETY ANALYSIS. LOG NO. 921 PROPOSED AMENDMENT NO. 155. PAGE 40

' t.

12. Existing Specification:

1, e

, N/A New Specification:

3.17.4

LIQUID EFFLUENT RADWASTE TREATMENT The LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the system shall.be used to reduce the quantity of radioactive materials in liquid effluents prior to their discharge to ensure that projected doses due to the liquid effluent beyond the Site Boundary for Liquid Effluents (see Figure 5.1-4) will not exceed the requirements of Specification 3.17.2.

Applicability At all times Action

a. With the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM ihoperable for more than.31 days or with radioactive liquid waste being discharged without treatment and in j excess of the above limits, prepare and submit to the Commission within_30 days pursuant to Specification 6.9.5 a Special Report which includes the following information:

1. Explanation of why liquid radwaste was being discharged without treatment, identification of

.any inoperable equipment or subsystems, and the reason for the inoperability,

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

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

Bases

'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

The installation of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM is not complete. This specification will become effective when the system.is declared operable.

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l FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 41 i

12. New Specification: (Cont.)

requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM were specified as the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

1 Discussion:

This new Tech Spec is added to reflect the addition of LIQUID EFFLUENT RADWASTE TRFATMENT SYSTEM (a sluiceable demineralized system) capable of polishing A & B RHUT contents prior to discharge to the retention basin on an as needed basis. This ensures that the contents of the A & B RHUT can be treated prior to being released to keep the liquid effluent within the do6e design objectives set forth in Section II.A of Appendix I of 10 CFR 50.

\

This new Tech Spec is added to be pursuant with the guidance of Standard Radiological Effluent Technical Specifications (RETS).

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FACILITY CHANGE SAFEIT ANALYSIS LOG NO. 921 j PROPOSED AMENDMENT NO. 155 PAGE 42 )

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13. Existing Specification: j 3.18 CASE 0US EFFLUENTS

/

3.18.1 DOSE RATE l

The dose rate at and beyond the site boundary due to radioactive materials released in gaseous effluents from j the site shall be limited to the following values: j

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

1

b. The dose rate limit for I-131, tritium, and for all radioactive materials in particulate form with half l lives greater than 8 days shall be 1500 mrem /yr to any l organ.

Applicability At all times Action g With the dose rate (s) exceeding the above limits, decrease l

- the release rate as soon as practicable to comply with the limit (s) given in Specification 3.18.1.

1 Basea !

This specification is provided to ensure that the dose rate at any time at the site boundary (see Figure 3.18-1) from gaseous effluents 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 concentration of 10 CFR Part 20, Appendir B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual outside the restricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendir B, Table II of 10 CFR Part !

20 (10 CFR Part 20.106(b)). For individual who may at times be j within the site boundary, the occupancy of the individual will be j sufficiently low to compensate for any increase in the atmospheric }'

diffusion factor above that for the restricted area boundary. The specified release rate limits restrict at all times the corresponding gamma and beta dose rates above background to an individual at or beyond the restricted area boundary to 500 mrem /yr to the total body or to 3000 mrem /yr to the skin. These release rate limits also j restrict at all times the corresponding thyroid dose rate above {

background to a child via the inhalation pathway to less than or I equal to 1500 mrem /yr. !

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 43

13. New Specification:

3.18 CASEOUS EFFLUENTS _

3.18.1 DOSE RATE The dose rate due to radioactive materials released in gaseous effluents from the site to areas at or beyond the Exclusion Area (see Figure 5.1-1) shall be limited-to the following values:

a. The dose rate limit for noble gases shall be less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin; and

b. The dose rate limit for Iodine-131, Iodine-133, tritium, and for all radioactive materials in particulate form with half lives greater than 8 days shall be less than or equal to 1500 mrem /yr to any organ.

Applicability At all times

\

Action With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the limit (s) given in Specification 3.18.1 and report the event in the next Semiannual Radioactive Effluent Report ' pursuant to Specification 6.9.2.3. ;

Bases This specification is provided to ensure that the dose rate from gaseous effluents at any time at the Exclusion Area Boundary (Figure 5.1-1) will be within the annual dose limits of 10 CFR Part 20, Appendir B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted j area to annual average concentrations exceeding the limits specified '

in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)(1)) . For individuals who may at times be within the .

Exclusion Area Boundary, the occupancy of the individual will be l sufficiently low to compensate for,any increase in the atmospheric diffusion factor above that for the site boundary.- The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the Exclusion Area Boundary to less than or equal to 500 mrem /yr to

p FACILITY CHANGE'SAFEff ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 44 13.' New' Specification: (Cont.)

the total body or to less than or equal to 3000'eres/yr to the skin.

These release rate limits also restrict, at all times, the

. corresponding thyroid dose rate above background to an infant via the inhalation pathway to less than or equal to 1500 mrea/yr.

Discussion:

Adoption to the Standard RETS is .lso incorporated in this change. The infant inhalation pathway has been identified as the limiting pathway for thyroid dose.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 45

14. Existing Specification:

3.18.2 NOBLE GASES The air dose at and beyond the site boundary due to noble gases released in gaseous effluents shall be limited to the following:

a. During any calendar quarter, to 5 mrad for gamma radiation and 10 mrad for beta radiation,

b. During 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, prepare and submit to the Commission within 30 days a Special Report. This Report will identify the cause(s) for exceeding the limit (s) and define the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

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

I FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED ' AMENDIGNT NO.155 PAGE 46

14. Existing Specification: (Cont.)

Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at or beyond the restricted area boundary (see Figure 3.18-1) will be i based on the historical average atmospheric conditions.

New Specification:

3.18.2 DOSE-NOBLE GASES The air dose due to noble gases released in gaseous effluents to areas at or beyond the Site Boundary for Caseous Effluents (see Figure 5.1-3) shall be limited to the following:

a. During any calendar quarter, to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation; and,

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

i Applicability At all times Action

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report pursuant to Specification 6.9.5. This Report will identify the cause(s) for exceeding the limit (s) and define the corrective action (s) to be taken to reduce the release of radioactive noble gases in gaseous effluents and the proposed corrective action (s) to be taken to assure that subsequent releases will be in compliance with the above annual limits.

Bases This , specification is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix 1. The ACTION statements provide the

(

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 47

14. New Specification: (Cont.)

required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirement implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual erposure of an individual through the appropriate pathways is unlikely to be I

substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion af Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining that the air doses at the Site Boundary,for Gaseous Effluents (Figure 5.1-3) are based upon th,e historical average atmospheric conditions.

Discussion:

This spec has been revised following the guidance in standard RETS.

FACILITY CHANGE SAFEIT ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO.155 PAGE 48

15. Existing Specification:

3.18.3 10 DINE-131, TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM Ihe dose or dose commitment to a member of the public from I-131, from tritium, and from radionuclides in particulate form with half-lives greater than eight days in gaseous effluents released at and beyond the site boundary shall 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 With the calculated dose or dose commitment from the release of I-131, tritium, and radionuclides in particulate from with half-lives greater than eight days in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report.

This Report will identify the cause(s) for exceeding the 0 limit and define the corrective actions to be taken to

. reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

Bases 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. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix 1 to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on modele and data cuch that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the restricted area boundary. The ODCM calculational methods for calculating the doses

4 l

l FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 j PROPOSED AMENDMENT NO. 155 PAGE 49 l

15. Existing Specification: (Cont.)

due to the actual release rates of the subject materials are required to be consistent with the Methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light Water-cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions.

The release rate specifications for radioiodines and particulate are dependent on the existing radionuclides pathways to man, beyond the site boundary. The pathways which were 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 erposure of man.

New Specification:

3.18.3 DOSE-IODINE-131, IODINE-133, TRITIUM AND RADIOACTIVE MATERIAL IN PARTICULATE FORM The dose or dose commitment to a MAXIMUM HYPOTHETICAL <

INDIVIDUAL from Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form with half-lives j greater than eight days in gaseous effluent released to areas at or beyond the Site Boundary for Gaseous Effluents (see Figure 5.1-3) shall be limited to the following:

a. During any calendar quarter to less than or equal to 7.5 mrem to any organ; and,

b. During any calendar year, to less than or equal to 15 mrem to any organ.

Applicability At all times Action With the calculated dose or dose commitment from the release of Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form with half-lives greater than

FACIIITY CHANGE SAFETY AFJ. LYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 50

15. New Specification: (Cont.)

eight days in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report pursuant to Specification 6.9.5. This Report will identify the cause(s) for exct+ ding the limit and defines the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent release will be in compliance with the above annual limits.

Bases This specifications 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. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance \ requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

The ODCM calculational methods for calculating the. doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for estimating doses based upon the historical average atmospheric conditions.

The release rate specifications for radioiodines and radioactive materials in particulate form are dependent on the existing radioneclide pathways to man in areas at or beyond the Site Boundary For Gaseous Effluents (Figstre 5.1-3). The pathways which were 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

t FACILITY CHANGE SAFEIT ANALYSIS LOG No. 921 PROPOSED AMENDMENT No. 155 PAGE 51

15. New Specification: (Cont.)

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.

Discussion l l

The changes are in conformance with the Standard RETS based on the dose to i the Maximum Hypothetical Individual. Also, clarification is made as to the site boundary for gaseous effluents.

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FACILITY CHANGE SAFETY ANALY5IS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 52

16. Existing Specification:

3.19 CASEOUS RADWASTE TREATM';23 The gaseous radwaste treatment system and the ventilation exhaust 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 noble gas releases e e and beyond the site boundary (see Figure 3.18-1), would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation over 31 days. The ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site to areas at or beyond the site boundary would exceed 0.3 mrem to any organ over 31 days.

Applicability When Gaseous Radwnste Treatment System and/or Ventilation Exhaust Treatment System are not being used.

\

Action

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

1. Explanation of why gaseous radwaste was being discharged without treatment, identification of the equipment or subsystems not OPERABLE and the reasons for inoperability.

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

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

Bases The OPERABILITY of the gaseous radwaste treatment system and the ventilation exhaust treatment system ensures that the systems will be available or 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

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I FACILITY CHANGE SAFETY ANALYSIS IDG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 53

16. Existing Specification: (Cont.)

effluents will be kept "as low as is reasonably achievable." The specification implements the requirements of 10 CFR Part 50.36a, ,

General Design Criterion 60 of Appendix A to 10 CFR Part 50 and !

design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the !

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.

New Specification:

3.18.4 GASEOUS RADWASTE TREATMENT The Waste Gas System and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of these systems shall be used to reduce radioactive materials in gaseous waste prior to their discharge such that projected gaseous effluent to areas at and beyond the Site Boundary for Gaseous Effluents (see Figure 5.1-3) are within the requirements of Specifications 3.18.2 and 3.18.3.

s Applicability At all times Action

a. With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, a Special Report pursuant to Specification 6.9.5 which includes the following information:

1. Explanation of why gaseous radwaste was being discharged without treatment, identification of the equipment or subsystems not OPERABLE and the reason for inoperability.

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

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

a recurrence, j i

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l FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 54

16. New Specification: (Cont.)

i Bases The OPERABILITY of the Waste Gas System and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the j 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 q Criterion 60 of Appendix I to 10 CFR part 50, and the design 1 objectives given in Section II.D of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the -

systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR part 50, for gaseous effluents.

Discussion:

The changes provide the correct name for the Waste Gas System and provide '

a reasonable basis for operability. Additionally, the changes are pursuant with the Standard RETS.

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l FACILITY CHANGE SAFETY ANALYSIS LOG-No. 921 i PROPOSED AMENDMENT NO. 155 PAGE 55 I

17 Existing Specification:

3.20- CAS STORAGE TANKS The quantity of' radioactivity contained in each waste gas decay tank shall be limited to 135,000 curies of noble gases (considered as Xe-133).

Applicability At all times Action When the reactor coolant system activity reaches the limit of Technical Specification 3.1.4, sample the online waste gas decay tank daily to ensure that the limit of 135,000 curies equivalent Xe-133 is not exceeded.

Bases i

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 exclusion area boundary will be exceed 500 mrem. This is consistent with Standard Review Plan 15.7.1,

" Waste Gas System Failure." .

Potential atmospheric releases from a waste gas decay tank are evaluated assuming design coolant activities (see page 14D-25 Vol. VI FSAR). Based on primary coolant activity as shown in Table 14D-7, the decay tank is assumed to hold the activity associated with the off-gas from one reactor coolant system degassing with no credit taken for decay.

Calculation of the limiting decay tank activity based on the coolant activity limit of Technical Specification 3.1.4 yields a maximum decay tank inventory of 98,414 Ci (Ref. FSAR Table 14D-23). In order )

for the decay tank inventory to reach the limiting condition for j operation, coolant activity would have to exceed the Technical i Specification 3.1.4 limit on coolant activity and this would require a reactor shutdown, thus preventing a further increase in gaseous activity.

1 Therefore, it is conservative to require that the online waste gas decay tank be sampled daily upon reaching the cooling limiting activity value (43/E) to ensure the'135,000 curies equivalent Xe-133 is not exceeded. Once the coolant is below the limiting activity, there is no requirement to sample waste gas decay tanks except for !

discharging. l 1

_ . _ . _ _ . . _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _J

FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 56

17. New Specification:

3.18 5 GAS STORAGE TANKS The quantity of radioactivity contained in each waste gas decay tank shall be limited to less than or equal to 135,000 curies of noble gases (considered as Xe-133).

Applicability At all times Action

a. With the quantity of radioactive material in any waste 3as decay tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.2.3.

Bases g Restricting the quantity of radioactivity contained in each waste gas d decay 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 Exclusion Area Boundary (see Figure 5.1-1) will not exceed 500 mrem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure."

Potential atmospheric releases from a waste gas decay tank are evaluated assuming design coolant activities (see page 14D-25 Vol. VI FSAR). Based on primary coolant activity as shown in Table 14D-7, the decay tank is assumed to hold the activity associated with the off-gas from one reactor coolant system degassing with no credit taken for decay.

Calculation of the limiting decay tank activity based on the coolant activity limit of Technical Specification 3.1.4 yields a maximum decay tank inventory of 98,414 Ci (Ref. FSAR Table 14D-23). In order for the decay tank inventory to reach the limiting condition for operation, coolant activity would have to exceed the Technical Specification 3.1.4 limit on coolant activity and this would require a reactor shutdown, thus preventing a further increase in gaseous activity.

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l- FACILITY CHANGE SAFETY ANALYSIS TOG NO. 921 PROPOSED AMENDMENT No. 155 . PAGE 57

17. New Specification: (Cont.)

Therefore, it is conservative to require that the online waste gas decay tank be sampled daily upon reaching the reactor coolant system limiting activity value (43/E) to ensure the 135,000 curies equivalent Xe-133 is not exceeded. Once the coolant is below the limiting activity, there is no requirement to sample waste gas decay' '

tanks except for discharging.

Discussion:

The Technical Specifications were renumbered from 3.20 to 3.19.5 and revised in accordance -with the guidance provided in the Standard RETS.

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9 FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 58

18. Existing Specification:

3.21 SOLID RADIOACTIVE WASTES The solid radwaste systems shall be used.in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial requirements.

Applicability At all times Action i

With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

Bases The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever radwastes require processing and packaging prior to being shipped offsite. This specification implements the r' requirements of 10 CFR 50.36a and General Design Criterion 60 of Appendix A to 10 CFR 50. The process parameters used in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to' waste type, waste pH, waste / solidification agent / catalyst ratios, waste oil content, vaste principal chemical constituents, mixing and curing times.

I New Specification: i i

3.21 SOLID RADIOACTIVE WASTES The solid radwaste systems shall be OPERABLE and used in accordance with a PROCESS CONTROL PROGRAM for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements ,of 10 CFR 20 and 10 CFR 71 prior to shipment of radioactive wastes from the site.

Applicability At all times Action

a. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 .

PROPOSED AMENDMENT NO. 155 PAGE 59

18. New Specification: (Cont.)

b. With the solid radwaste system inoperable for more than 31 days, prepare and submit to the commission within 30 days pursuant to Specification 6.9.5 a special Report which includes the following information:

1. Identification of the inoperable equipment or sub-systems and the reason for inoperability.

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

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

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

Bases

\

The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR 50.36a and General Design Criterion 60 of Appendix A to 10 CFR 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, vaste oil content, waste principal chemical constituents, mixing and curing times.

Discussion:

The action statement added the reporting requirement in accordance with the guidance in the Standard RETS.

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I JACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 60 4

I

19. Existing Specification:

3.22 RADIOLOGICAL ENVIRONMENTAL MONITORING j The Radiological Environmental Monitoring Program shall be conducted as specified in Table 3.22-1.

Applicability At all times Action j

a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 3.22-1, 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. (Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, or seasonal unavailability, or to m'a lfunction of automatic sampling equipment. If the latter, efforts 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 exceeding the reporting level of Table 3.22-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days after the level of radioactivity has been determined, a Special Report pursuant to Specification 6.9.5 which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits to be exceeded. This report will define corrective actions to reduce emissions such that potential annual exposures will meet the Specifications 3.17.2, 3.18.2, and 3.18.3. The exceeding of Table 3.22-2 levels may result from more than one radionuclides in the sampling medium if:

Concentration (1) + Concentration (2) + >1.0

~

reporting level (1) reporting level (2)

' Dose calculations will include all measured radionuclides of plant origin. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

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f FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 .PAGE 61

19. Existing Specification: (Cont.)

c. With milk or fresh leafy vegetation samples unavailable from any of the sample locations required by Table 3.22-1, prepare and submit to the Commission i within 30 days a Special Report which identifies the cause of the unavailability of samples and identifies locations for obtaining replacement samples. The locations from which samples were unavailable may then be deleted from Table 3.22-1 provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations,~ if available.

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

The specified monitoring program is in effect at this time. Program changes may be initiated based on operational experience, and changes in regional population or agricultural practices. The sample locations have been listed in the ODCM to retain flexibility for making changes as needed.

With no drinking water intakes downstream of the plant, surface water and runoff water samples do not have to meet drinking water requirements and sample frequencies.

New Specification:

3.22 RADIOLOGICAL ENVIRONMENTAL MONITORING The Radiological Environmental Monitoring Program shall be conducted as specified in Table 3.22-1.

Applicability At all times Action

a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 3.22-1, prepare and submit to the Commission, in the Annual l

d

FACILITI CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 62

19. New Specification: (Cont.)

Radiological Environmental Operating Report required by Specification 6.9.2.2, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. (Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, or seasonal unavailability.)

b. Witil the level of radioactivity in an environmental sampling medium exceeding the reporting level of Tabic 3.22-2 when averaged over any calendar quarter, prepare and submit to the commission within 30 days after the level of radioactivity has been determined, a Special Report pursuant to Specification 6.9.5 which includes an evaluation of any release conditions, environmental factors or other aspects which caused the reporting limits to be exceeded. This report will define corrective actions to reduce emissions such that potential exposures will meet Specification 3125. When more than one of the radionuclides in Table 3.22-2 are detected in the sampling medium, this I report shall be submitted if: i i

Concentration (1) + Concentration (2) >1.0 reporting level (1) reporting level (2)

When radionuclides other *han those in Table 3.22-2 are detected and are the result of plant effluents, ,

this report shall be submitted if the potential annual j dose to an individual is equal to or greater than the I calendar year limits of Specification 3.17.2, 3.18.2, '

and 3.18.3. This report is not required if the measures level of radioactivity was not the result of plant effluents; however, the condition shall be reported and described in the Annual Radiological Environmental Operating Report. 3

c. With milk or fresh leafy vegetation samples unavailable from any of the sample locations required by Table 3.22-1, prepare and submit to the Commission within 30 days a Special Report pursuant to Specification 6.9.5 which identifies the cause of the unavailability of samples and identifies locations for obtaining replacement samples. The locations from which samples were unavailable may then be deleted i

__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ a

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 63

19. New Specification: (Cont.) !

from Table 3.22-1 provided the locations from which the replacement samples were obtained are added to the Radiological Environmental Monitoring Program as replacement locations, if available.

Bases The Radiological Environmental monitoring Program required by this ;

specification provides measurements of radiation and of radioactive .

materials in those exposure pathways and for those radionuclides J which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and ODCM modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The s,pecified monitoring program is in effect at this time. Program changes may be initiated based on operational experience, and changes in regional population or agricultural practices. The sample locations have been listed in the RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

Manual to retain flexibility for making changes as needed.

1 The detection capabilities required in Table 4.26-1 are state-of-the-art for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirement of 40 CFR 141.

Discussion The text changes to the Action statements are editorial and clarification changes. Specifically,

1. " Environmental" inserted between " Radiological" and " monitoring" to differentiate between onsite protection and offsite environmental monitoring.

2. " Reporting" inserted between " exceeding the" and " level" in item b.

to clarify the content of Table 3.22-2.

The test changes to the Bases statements are similar to those in the Action statements. The sample locations are defined in the REMP manual and not the ODCM.

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i FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 63

19. New Specification: (Cont.)

from Table 3.22-1 provided the locations from which !

the replacement samples were obtained are added to the Radiological Environmental Monitoring Program as replacement locations, if available.

Bases The Radiological Environmental 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 implementsSection IV.B.2 of Appendix I to 10 CFR 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurcble concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and ODCM modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position, Revision 1, November 1979. sThe specified monitoring program is in effect at this time. Program changes may be initiated based on operational experience, and changes in regional population or agricultural

, practices. The sample locations have been listed in the RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) Manual to retain flexibility for making changes as needed.

The detection capabilities required in Table 4.26-1 are state-of-the-art for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirement of 40 CFR 141.

Discussio,n The text changes to the Action statements are editorial and clarification changes. Specifically,

1. " Environmental" inserted between " Radiological" and " monitoring" to differentiate between onsite protection and offsite environmental monitoring.

2. " Reporting" inserted between " exceeding the" and " level" in item b.

to clarify the content of Table 3.22-2.

i The test changes to the Bases statements are similar to those in the Action statements. The sample locations are defined in the REMP manual and not the ODCM.

i FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 64

19. New Specification: (Cont.)

The changes represent conformance with the Standard RETS to provide reporting levels for radionuclides concentrations in the environmental samples (Table 3.22-2) in order to appropriately identify when concentrations of radioactive materials and levels of radiation may be higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. In addition, the Radiological Environmental Monitoring Program (REMP) will account for all potential land, water usage, and food radiological exposure pathways that exist downstream from Rancho Seco. The sampling and collection frequency (Table 3.22-1) will allow determination of long-term buildup of concentrations of radionuclides in bottom sediment, doses due to ingesting aquatic foods (bottom feeding fish) and direct radiation from long-term buildup of radionuclides on land irrigated with contaminated water.

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20. Existing Specification: Page 65

. Table 3.22-1

~

RADIOLOG bL ENVIRONMENTAL 50EITORIEG G ER0' RAM - - -

Sampling and i Exposure Pathway Number of Collection Type and Frequency and/or SampPe Sampl es* Frequency of Analysis

1. AIRBORNE A. Radiofodine 8 Continuous oper- Radiofodine canis-and Parti- at.on of sampler ter. Analyze at culates collection as least once weekly required by dust for I-131.

loading but at least once per particulate week. sampler. Analyze for Gross . Beta radicactivi ty greater- than or equal to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following filter change. Perfor1s

\ gama isotopic

~

anal / sis on each sample where gross

betz activity is greater than 10 times the appro-priate control samples for the same sample period.

Perform gama iso-

. topic analysis on composite (by location) sample

. at least once per quarter. ,

2. DIRECT Greater than 40 At least once Gama dose. At RADIATION locations with 2 per quarter. least once per dosimeters at each quarter.

location.

Sample locations are shown in the 00CM.

20. Existing Specification: (Cont.) Page 66

~ ~~

~~ ^ Table 3.22-1 (Continued)

RADIOLOGIbLENVIRONMENTALMONITORINGPROGRAM Sampling and Exposure Pathway Number of Collection Type and Frequency and/or Sample Samples

Frequency of Analysis

3. WATERBORNE t

a. Surface 3 Grab sample Gross Beta and collected I-131 analysis of monthly , each suspended and ~;

dissolved fraction.

Tritium inalysis 'at

.least once per quarter.

b. Runoff 1 Grab sample Gross Beta and

. collected I-131 analysis of fortnightly. each suspended and

\ -

dissolved fraction.

Tritium analysis at

least once per quarter, plus genma isotopic analysis

. on dissolved and suspended frac ~

tions.

c. Mud and 2 At least once Gross Beta on Stit semi-annually. each sample.

One pint sampie of the top 3" of material

. 2 ft. from ,

shoreline.

Sample locations are shown in the 00CM.

_ . _ _ _ _ _ _ _ _ - J

.1 20.' Existing Specification: (Cont.) Page 67 1

i

~

, Table 3.22-1 (Continued) 1

, RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM - ;

Sampling and Exposure Pathway Number of Collection Type and Frequency and/or Sample _ Samples

Frequency of Analysis

4. INGESTION

a. Milk 4 At least once I-131 analysis of per fortnight each sample.**

when animals are on pasture; I at least once per month at .

other times.

b. Fish 1 At least semi- Gross Beta minus annually. One K-40 analysis on i sample of each edible portion of l each sample.**

of several species as

\

shown in the ODCM.

l

c. Food 4 At time of har- Gross Beta minus vest. One sam- K-40 analysis on edible portion of pie of each of the several ,

each sampl e.**'

classes of food products as shown in the 00CM.

1

Sample locations are shown in the 00CM.

- Gansna Isotopic Analysis when Table 3.22-2 levels are exceeded.

f

- _---_\--_ -______-.-_____.-__A__.-__-

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4 20. New Specification:- Page 68 Table 3.22-1 ;

-RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM-

' Sampling and, Exposure Pathway Number of. Collection Type and Frequency

.and/or Sample ' Samples

Frequency of Analysis

1. AIRBORNE ~[

A. Radiciodine 8 Continuous oper Radioiodine canis-and Parti-ation'of sampler ter. Analyze at=

culates with sample least once weekly collection as for I-131. t

' required by dust loading but at Particulate least once per sampler. Analyze week. f'or Gross Beta radioactivity at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months t-following filter change. Perform

- gamma isotopic.

.g analysis on each:

sample where gross ,

- 9 beta activity is greater than 10 times the yearly s

mean of control: '

samples for the same sample period.

Perform gamma iso-topic analysis on composite (by l'ocation) for particulate filters sample at least once per. quarter.

2. DIRECT Greater than 40 At least once Gamma dose. At RADIATION locations with 2 per quarter. least once per dosimeters at each quarter.

location.

Sample locations are shown in the REMP MANUAL.

_y - -

- . )

l

20. New Specification: (Cont.) Page 69 Table 3.22-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i

Sampling and Exposure Pathway Number of Collectior, Type and Frequency and/or Sample Samples

Frequency - of Analysis

3. WATERBORNE ,

4 Surface Composite Gamma isotopic

a. 1 sample collected and tritium monthly ** analysis of each

' composite.

3 Grab sample Gamma isotopic and collected tritium analysis of each sample.

monthly.

b. Runoff 1 Grab sample Gamma isotopic and collected tritium analysis of fortnightly. each sample.

c. Ground 2 At least once Gamma isotopic, per quarter. and tritium J

analysis of each sample.

At least once Gamma Isotopic

d. Mud and 2 Silt semi-annually. analysis of One pint samgle each sample.

of the top 3 of material 2 ft. from shoreline.

Sample locations are shown in the REMP MANUAL.

- Applicable when sampler is declared operational.

M e

20. New Specification: (Cont.) Page 70, Table 3.22-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Sampling and .

Exposure Pathway Number of Collection Type and Frequency

. .and/or Sample Samples

Frequency of Analysis ,

4. INGESTION

a. Milk 4 At least weekly Gamma isotopic when animals analysis and are on pasture; I-131 analysis of at least once each sample.

per month at other times.

. Fish e.id 3 At least Gamma isotopic Inverte- quarterly. One analysis on edible brates sample of each portion of each species as sample.

listed in the

\ REMP MANUAL.

c. Food 4 At time of har- Gamma isotopic, vest. One. sam- analysis on pie of each of edible portion _

the several of each sample.

classes of food products as shown in the REMP MANUAL.

Technical Specification sample locations are identified in the REMP MANU.AL.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 71-

- 20. New Specification: (Cont.)

l Discussion The changes made in Table 3.22-1 support additional sampling and '

collection frequency of the liquid effluent pathway documented in the REMP manual. The NRC considered that the District was deficient in this aspect of the Rancho Seco radiological effluent monitoring program. Also, the

. Table.3.22-1 changes reflect conformance to the Standard RETS by the addition of a monthly composite sample and the deletion of Gross Beta and 1-131 analysis for the waterborne surface exposure pathway.

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FACILITY CHANGE SAFE 1T ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 74

21. New Specification: (Cont.)

Discussion l Table 3.22-2 values represent a reporting level of radioactive )

concentrations in environmental samples and not a regulatory limit. The l requirements cre a report that evaluates the effluent observations against the EPA regulations documented in 40 CFR 190 (Tech Spec 3.25).

Additional reporting levels are included for cesium and icdine (I-131) anc?,ysis in fish and food products which are in the liquid effluent pathway. This is in conformance with 'he Standard RETS.

\

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FACILITY CHANGE SAFETY ANALYSIS . LOG NO. 921 FROPOSED AMENDMENT NO. 155 PAGE 75

22. Existing Specification:

3.23 IAND USE CENSUS A land use census shall be conducted annually and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden

of greater than 500 square feet producing fresh leafy vegetables in each of the 16 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 4.22.3, identify the new locations in the next Semiannual Radioactive Effluent Release Report.

b. W4th a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same erposure pathway) 20% greater than at a location d from which samples are currently being obtained in accordance with Specification 3.22, add the new location (s) to the radiological environment monitoring program within 30 days. The sampling location (s),

excluding the control station location, having the lowest calculated dose or dose commitment (s) (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. Identify the new location (s) in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting i the new location (s).

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

Bases i

This specification is provided to ensure that changes in the use of areas at and beyond the site boundary 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 i Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census

FACILITT CHANGE SAFEIT ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 75

22. Existing Specification: (Cont.)

]

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/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this i minimum garden size, the following assumptions were used: (1) that 20% of the garden was used for growing broad leaf vegetation (i.e.,

similar to lettuce and cabbage); and (2) a vegetation yield of 2 kg/ square meter.

New Specification:

3.23 LAND USE CENSUS A land use census shall be conducted annually and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden

of greater than 500 square feet producing fresh leafy vegetation in each of the 16 meterological sectors within a distance of five miles.

The La d Use Census shall also include information relevant to the liquid effluent pathway and gaseous effluent pathway such that the OFFSITE DOSE CALCULATION MANUAL (ODCM) and the RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL (REMP) can be kept current with the existing environmental and societal uses surrounding Rancho Seco.

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 Specifications 4.21.2, and 4.22.3, identify the new locations in the next Annual Radiological Environmental Operating Report.

b. With a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 3.22, add the new location (s) to the Radiological Environmental Monitoring Program within 30 days or submit a Special

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 77

22. New Specification: (Cont.)

Report to the Commission pursuant to Specification 6.9.5 that identifies the cause(s) for exceeding these requirements and the proposed corrective actions for precluding recurrence. The sampling location (s),

excluding the control station location, having the lowest calculated dose or dose commitment (s) (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. Identify the new location (s) in the next Annual Radiological Environmental Operating Reporting and also include in the report a revised figure (s) and table for the REMP manual reflecting the new location (s).

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

Bases

\

This specification is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the Radiological Environmental Monitoring Program and the ODCM 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 erposure 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/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used: (1) that 20 percent of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage); and (2) a vegetation yield of 2 kg/ square meter.

In addition, by gathering information of the liquid effluent pathway and the gaseous effluent pathway, the census will ensure that proper radiological environmental monitoring and radioactive effluent controls are in place for the adequate ptotection of the haalth and safety of the general public.

Discussion The Rancho Seco REMP, utilizing the guidance of NUREG-0472, provide for an annual land use census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modification to the

l FACILITY CHANGE SAFETY ANALYSIS LOG No. 921' PROPOSED AMENDIGNT No.155 PAGE 78 2.2, New Specification: (Cont.)

l monitoring program'are made if required by the results of the census. The changes here will include an addition of liquid pathway surveillance so that existing environmental and societal uses of land surrounding Rancho Seco can be kept current. Identification of' gardens in the summer, rather !

'than the middle of winter will be included in the census to assure a more l realistic' sampling of gardens. In addition, liquid and gaseous pathways 1 are identified and reportable as land use. census dose results which will be included in the Annual Radiological Environmental Operating Report.

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FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 79 1

23. Existing Specification:

l 3.25 FUEL CYCLE DOSE The annual dose or dose commitment to a member of the public due to releases of radioactivity and radiation from uranium fuel cycle sources is limited to _25 mrem to the ,

) total body or any organ (except the thyroid, which is limited to _75 mrem).

Applicability At all times Action With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.17.2.a, 3.17.2.b, 3.18.1.b, 3.18.2.a, 3.18.2,b, 3.18.3.a or 3.18.3.b, calculations should be made to determine whether the above limits of Specification 3.25 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405(c), shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations. If the estimated dose (s) exceed the above limits, and if the release condition resulting in violation of 40 CFR Part 190 hastnot already been corrected, the Special Report shall inciude a request for a variance in accordance with the profision of 40 CFR Part 190. Submittal of the report is con'.jidered a timely request, and a variance is granted untrl staff action on the request is complete.

Bases This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Rep rt whenever the calculated doses from the plant r adioactivo effluents exceed twice the design objective doses of

!ppendix I. For the Rancho Seco site it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits

-.-__._.____.___4 _0______.______

l FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 i PROPOSED AMENDMENT NO. 155 PAGE 80 i

'23. E11 sting Specification: (Cont.)

of 40 CFR 190 if the plant remaint within the reporting requirement level. The Special Report will dese. ribe a course of action which should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the doce commitment to' the MEMBER OF THE PUBLIC from other ure.nium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of.

5 miles must be considered. If the dose to,any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.1 until -

NRC staff action is completed. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle.

New Specification:

3.25 FUEL CYCLE DOSE The dos \e or dose commitment to any real MEMBER OF THE PUBLIC due to releases of radioactive material in gaseous and liquid effluents and to direct radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months.

Applicability At all times Action

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.17.2.a, 3.17.2.b, 3.18.2.a, 3.18.2.b, 3.18.3.a, or 3.18.3.b, or exceeding the reporting levels of Table 3.22-2, calculations shall be made including direct radiation contributions (including outside storage tanks, etc.)

to determine whether the above limits of Specification 3.25 have been exceeded.

b. If the above limits have been exceeded, prepare and l submit to the Commission within 30 days, a Special Report pursuant to Specification 6.9.5 that defines the corrective action to be taken to reduce subsequent

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 81 i

23. New Specification: (Cont.)

releases to prevent recurrence of exceeding the above i limits and includes the schedule for achieving i conformance with the above limits. This Special Report, as defined in 10 CFR pcrt 20.405(c), shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, over 12 consecutive months that includes the release (s) covered by this report. It shall also describe levels of radiation and 1 concentrations of radioactive material involved, and the cause of the exposure levels or concentrations.

c. If the estimated dose (s) exceed the above limits, and if the release condition resulting in the violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provision of 40 CFR Part 190.

Submittal of the report is considered a timely request, and a variance is granted until staff action on the request in complete.

Bases This specification is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the numerical guides for design objective doses of Appendix I or the reporting levels for the Radiological Environmental Monitoring Program. For the Rancho Seco site it is unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the plant remains within twice the numerical guides for design objectives of 10 CFR 50 Appendix I and if direct radiation (outside storage tanks, etc.) is kept small. The Special Report will describe a course of action which should result in the limitation of the dose to a MEMBER OF THE PUBLIC for 12 consecutive months to within the 40 CFR 190 limits. For the purpose of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any MEMBER OF THE PUBLIC is evaluated to l exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in

_ _ _ = _ ,

6' FACILITY CHANGE SAFETY ANALYSIS - LOG NO. 921 PROPOSED AMENDMENT NO.155 PAGE 82

23. New Specification: (Cont.)

accordance with the provisions _of 40 CFR 190 is considered to be a timely' request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a MEMBER OF THF PUBLIC during any period in which he/she is engaged in carrying out any operation which is part of the uranium fuel cycle.

Discussion Changes represent conformance to the Standard RETS, by the deletion of compliance to 10 CFR 20.

\

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 83

24. Existing Specification:

3.26 INTERLABORATORY COMPARISON PROGRAM The contractor performing the analysis of radiological environmental program samples for radioactive materials shall participate in an Interlaboratory Comparison program approved by the Commission.

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

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

New Specification:

3.26 INTERLABORATORY COMPARISON PROGRAM The contractor performing the analysis of radiological environmental monitoring samples for radioactive materials shall participate in an Interlaboratory Comparison Program approved by the Commission.

Applicability ' At all tines Action With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.2.2.

i FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 84

24. New Specification: (Cont.)

)

Bases j The requirement for participation in an Interlaboratory Comparison ]

Program is provided to ensure that independent checks on the j precision and accuracy of the measurements of sdioactive material in .

environmental sample matrices are performed as cart of the quality assurance program for environmental monitoring in order to demonstrate that the results 'are reasonably valid for the purposes of-Section IV.B.2 of Appendix I to 10 CFR 50.

Discussion:

This specification has been revised to clarify and to reference the specification requiring submittal of the Annual Radiological Environmental

' Operating Report.

I

\

25. Existing Specification: Page 85 Taole 4.1-1 (t.ontinueos INSTRUMENT SudVEiLi.AnCE Rt00!REidhTS Channel Description Check Test Caliorate Remarks

44. deactor out tuing urain accumulation tank level NA NA d

43. Incore neutron detectors 'M(1) WA NA (1) Check functioning, including functioning of computer read-

, out and/or recorder reacout.

44 a. Process and area radi-ation monitoring system W H Q

b. Containment Area Monitors W NA R

45. Emergency plant radiatfori Instruments M(i) NA R (1) tlattery check

46. Environmental air monitors M(1) NA R (1) Check functioning

47. Strong motion accelerometer y( 1) NA R (1) dattery check

48. Auxiliary Feedwater ,

Start Circuit '

a. Phase lacalance/Under-puwer RCP S WA d

b. Low Main Feedwater ;

3 Pressure dA rt R

\ -

49. Pressurizer Water Level M NA R I

50. Auxiliary Feedwater flow Rate N NA R

51. Reactor Coulant b/ stem Suo-cooling Margin Monitor M NA R

52. EHOV Pu.,er Position Indicator (Priaary Detector) A NA 4

53. EMOV Position Incicator l (Wackup DetectorJ A NA R T/C or Acoustic 54 EHOV Block Valve Position Indicator r4 NA R

55. Safety Valve Position In-dicator (Primary Detector) M NA R T/C

56. Sarety Valve Position In-01cator (Backup Detector)

Acoustic M l NA R g e

25. New Specification: Page 86 Table 4.1-1 (Continued)

, INSTRUMENT SURVC1LLANCE REQUIREMENTS.

Chect Test Calibrate Rematis l Channel Description

42. Reactor Building drain NA NA R l

acctsuulation tant level M(1) NA NA (1) Check functioning.

43. Incore neutron detectors including functioning of computer readout and/or recorder readout.

44. a. Process radiation R monitoring systen W Q

b. Area radiation monitoring systen W H Q W NA R

c. Contatraent Area Monitors

45. Emergency plant radiation (1) 8attery check M(1) NA R Instrveents M(1) NA R (1) Chect functioning

46. Envircreental air monitors Q(1) NA R (1) Battery check

47. Strong motion accelerometer

48. Auxiitary Feedwater Start Circuit

a. Phase imbalance /Under-power RCP S MA R

b. Low Main Feedwater \

Pressure NA M R M NA R

49. Pressurizer Water Level '

50. Auxiliary Feedwater Flow M NA R Rate

51. Reactor Coolant System .

Subcooling Margin Monitor M NA R 52 OMOV Power Position Indicator (Primary Detector) H HA R

53. EMOV Position Indicator (Backup Detector)

T/C or Acoustic M NA R

54. EMOY Block Yalve Position Indicator M NA R

$5. Safety Yalve Position Indicator (Primary Detector)

M NA R T/C

56. Safety Valve Position Indicator (Backup Detecto -) .

Acoustic M NA R Discussion: ,

Revise in accordance with the guidance of the Standard RETS.

I

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 87

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26. Existing Specification:

4.19 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION Surveillance Requirements i

l The maximum setpoint shall be determined in accordance with 1 procedures as described in the ODCM and shall be recorded on the release permits.

Each radioactive liquid effluent monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the INSTRUMENT CHANNEL CHECK, SOURCE CHECK, INSTRUMENT CHANNEL CALIBRATION, AND CHANNEL TEST at the frequencies shown in Table 4.19-1.

Records shall be maintained in the Process Standards of all radioactive liquid effluent monitoring instrumentation alarm / trip setpoints. Maximum setpoints and setpoint calculations shall be available for review to ensure that the limits of Specification 3.17 are met.

Bases

\

The radioactive liquid effluent instrumentation is provided to

, monitor and control, as applicable, the releases of radioactive I

materials in liquid effluents during actual releases. The alarm / trip setpoints for these instruments shall be calculated in accordance wit methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this inst.rumentation is consistent with the requirements of General Design criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

New Specification:

4.19 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Surveillance Requirements The maximum setpoint shall be determined in accordance with methodology as described in the Offsite Dose Calculation Manual (0DCM) and shall be recorded on the release permits.

Each radioactive liquid effluent monitoring instrumentation shall be demonstrated OPERABLE by performance of the INSTRUMENT CHANNEL CHECK, SOURCE CHECK, INSTRUMENT CHANNEL CALIBRATION, AND CHANNEL TEST at the frequencies shown in Table 4.19.1.

FACILITY CHANGE SAFEIT ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO.'155 PAGE 88

26. New Specification: (Cont.)

Records shall be maintained in the Process Standards of all radioactive liquid effluent monitoring instrumentation alarm / trip setpoints. Maximum setpoints and setpoint calculations shall be available for review to ensure that the limits of. Specification 3.17.1 are met.

Bases The radioactive liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of i radioactive materials in liquid effluents during actual or potential release of radioactive liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the methodology contained in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.106. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

Discussion The changes represent conformance to the Standard Radiological Environmental Technical Specifications (RETS). Also reflected is that only the ODCM contains the methodology to ensure that the setpoints are adequate for the alarm / trip to occur prior to exceeding the limits of 10 CFR 20.106.

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27. Existing Specification: -

Page 89 )

l Table 4.19-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Instrument Channel Source Channel Channel Instrument Check Check Calibration Test

1. Gross Beta or Gamma Radioactivity Monitors Providing Alarm and Automatic Isolation

a. Regenerant Hold-Up Tank Discharge Line (1) (5) (2) (3)

Monitor D M R Q

2. Flow Rate Monitors (4)

a. Waste Water Flow D NA R NA

Table Notation

- \ _

(1) During releases via this pathway, a check shall be performed at least

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

(2) The Instrument Channel Calibration for radioactivity measurement instrumentation shall be performed using one or more reference standards.

(3) The Channel Test shall also demonstrate that autoaatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist: l

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

b. Circuit failure.

c. Instrument indicates a downscale failure.

d. Instrument controls not set in operate mode.

(4) The Instrument Channel Check shall consist of verifying indication of flow during periods of release. The Instrument Channel Check shall be made at least once daily on any day on which batch releases are made.

(5) During periods of known activity in the regenerant tank, perform a source check daily during releases via this pathway.

- _ _ - - - _ _ _ - - _ _ _ _ _ - . _ _ _ - - _ - _ _ _ - _ _ _ _ _ _ _ - - _ __-.__a_--__-_____--._-----______------,----___---_--_____ _ _ - - - - - -

27, New Specif'ication: Page 90 Table 4.19-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument ..

Instrument !

Channel Source Channel Channel Instrument Check- Check Calibration -Test

1. Gross Radioactivity Monitors Providing Alarm and Automatic Isolation

a. Retention Basin D P R(2) g(1)

Effuent Discharge Monitor

2. Flow Monitors

a. Regenerant Hold-up D(3) NA R Q:

Tank Discharge Line Total Flow

\

b. Waste Water Flow D(3) NA R Q Rato .

TABLE NOTATION (1) The CHANNEL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the !

following conditions exists:

Instrument indicates measured levelt above the alarm / trip setpoint.

1.

2. Circuit failure. ,

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

(2) The INSTRUMENT CHANNEL CALIBRATION shall be performed using one or more reference stai1dards (3) The INSTRUMENT CHANNEL CHECK shall consist of verifying indication of flow during periods of release. INSTRUMENT CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 l~ PROPOSED AMENDMENT NO. 155 .PAGE 91

27. New Specification: (Cont.)'

l Discussion New monitors are added'on the Retention Basin Effluent, which is now the 3 i' current Rancho Seco environmental control point. The additions to the table notation are pursuant with Standard RETS. j 1

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28. Existing Specification

Page 92 4.20 RADI0 ACTIVE GASEOUS EFFLUENT MONITCRING INSTRUMENTATION Surveillance Requirements The maximum setpoints shall be determined in accordance with procedures as described in the ODCM and shall be recorded on release permits.

Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the INSTRUMENT CHANNEL CHECX, SOURCE CHECX, INSTRUMENT CHANNEL CALIBRATION, AND CHANNEL TEST at the frequencies shown in Table 4.20-1.

Records shall be maintained in the Process Standards of all radioactive gaseous effluent monitoring instrumentation alarm / trip setpoints. Maximum setpoints and setpoint calculations shall be available for review to ensure that the limits of Specification 3.18 are met.

Bases The radioactive gaseous effluent instrumentation is provided to monitor ano control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm / trip setpoints for these instruments shalil be calculated in accordance with the methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements and General Design Criteria 60, 63, anc 64 of Appendix A to 10 CFR Part 50.

The flow rates in the Reactor Building Purge Vent, Auxiliary Suilding Stack and Radwaste Service Area Vent are constant as they use single speed fans.

The Reactor Building Purge Vent has two different flow rates, winter und summer, however administrative controls assure using the correct flow rate where applicable. The actual flow rate of the ventilation systems are periodically determined by surveillance procedures. The flow rate measurement devices are used only as flow indicating devices anc not for actual measurement of flow rate. Also, as these flow rate devices must be removed from the ventilation system for the channel test, and in acdition transported to the manufacturer for calibration, the frequencies have been set as shown in Table 4.20-1.

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28. New Specification: Page 93 4.20 RADIOACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION Surveillance Requirements The maximum setpoints shall be-determined in accordance with methodology described in the OFFSITE DOSE CALCULATION MANUAL (0DCM) and shall be recorded on release permits.

Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the INSTRUMENT CHANNEL CHECK, SOURCE CHECK, INSTRUMENT CHANNEL CALIBRATION, AND CHANNEL TEST at the frequencies shown in Table 4.20-1.

j Records shall be maintained in the Process Standards of all radioactive gaseous effluent monitoring instrumentation alarm / trip setpoints. Maximum setpoints and setpoint calculations shall be available for review to ensure that the limits of Specification 3.18.1 are met. l Bases i The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of radioactive gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the methodology contained in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.106.

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

The flow rates in the Auxiliary Building Stack and Auxiliary Building Grade ;

Level Vent are constant as they use single speed fans. The Reactor Building ,

Purge Vent has three different flow rates, winter, summer and mini-purge, however administrative controls assure using the correct flov rate where i applicable. The actual flow rate of the ventilation systen, are periodically determined by surveillance procedures. The flow rate measurement devices are used only as flow indicating devices and not for actual measurement of flow rate. Also, as these flow rate devices must be removed from the ventilation i system for the channel. test, and in addition transported to the manufacturer I for calibration, the frequencies have been set as shown in Table 4.20-1.

Discussion:

Clarification is made that provides that only the ODCM describes the method-ology to establish the setpoints to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR 20.106. The "Radwaste Service Area Vent" names has been editorially changed to " Auxiliary Building Grade Level Vent."

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29. Existing Specification: Page 94

Table 4.20-1 ,

RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENT 5 Instrument , Instrument Channel Source Channel Channel Check Check Calibration Test Instrument

1. Reactor Building Purge Vent

a. Noble Gas Activity Monitor Dill M Q(2) g(3) ,

Iodine Sampler NA NA

b. W NA

' c. Particulate NA Sampler W NA NA

d. System Effluent NA BY A Flow Rate Device W

e. Sampler Monitor Flow Rate NA BY A g

Measurement Device W

2. Auxiliary Building Stack

a. Noble Gas Activity Monitor D(1) M Q(2) g(3)

6. Iodine Sampler W NA NA NA

c. Particulate NA NA NA Sampler W

d. System Effluent Flow Rate Device

W NA BY A ;

e. Sampler Monitor Flow Rate NA BY A Measurement Device W

This flow rate device is not yet installed. This specification for this system will become effective when it is declared OPERABLE.

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29. Existing Specification: (Cont.) , 'Page 95 Table'4.20-1 ,

(Continued) ;

Instrument Instrument Channel Source Channel Channel Instrument Check Check Calibration Te st --

3. Radwaste Service Area * ~ ,

a. Noble Gas g(4)

' Activity Monitor DIl) M Q(2)

NA NA

b. Iodine Sampler W' NA

c. Particulate NA Sampler W NA NA

d. System Effluent BY A Flow Rate Device W NA

e. Sampler Monitor Flow. Rate NA SY A Measurement Device W

The Radwaste Service Area Monitoring System is not yet functional. The -

specification for this system wi+1 become effective when it is declared ~~

OPERABLE. ,

e-Table Notation (1) During releases via this pathway, 'a check shall be perfonned at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(2) The Instrument Channel Calibration for radioactivity measurement instrumentation shall be perfonned using one or more reference standards.

(3). The Channel Test shall also demonstrate that automatic tennination of this pathway and control room alarm annunciation occurs if any of the following conditions exist: -

a. Instrument indicates measured levels above the alann/ trip setpoint.

b. Circuit failure.

c. Instrument indicates a downscale failure.

d. Instrument controls not set in operate mode.

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1 29.ExistingSpecification: (Cont.) Page 96 ,

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. Table 4.20-1 (continued) j RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4 i

SURVEILLANCE REQUIREMENTS (4) The Channel Test shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

a. -Instrument indicates measured levels above tihe alann/ trip setpoint.

b. Circuit failure.

c. Instrument indicates a downscale failure.

<- d. Instrument controls not set in operate mode.

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N 29.-Nek Specification: Page 97 Table 4.20.-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION

, SURVEILLANGL REQUIREMEN15 , , ,

^

'. - . Instrument -

Instrunient Channel Source Chann&1 Chinnel-Instrument Check Check Calibration - Test

1. ' Reactor Building Purge Vent

a. Noble Gas Activity Monitor D M(.4 } ' R(3) .Q(1)-

b. . Iodine Sampler W NA NA NA

c. Particulate Sampler W NA NA NA d .' System Effluent Flow Rate Device D NA R Q(6)

e. Sampler Monitor Flow Rate '

Measurement Device O NA R Q Auxiliary Building 2.

Stack

a. Noble Gas Activity Monitor D(5) M R(3) Q(7)

b. Iodine Sampler W ,

NA NA NA

c. Particulate Sampler W NA NA NA

d. System Effluent Flow Rate Device D NA R Q(6)

e. Sampler Monitor Flow Rate Measurement Device D NA R Q i

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29. New Specification: (Cont.) Page 98 Table 4.20-1 (Continued)

Instrument Instrument Channel Source Channel Channel Check Check Calibration Test Instrument '

3. Auxiliary Building .

Grade Level Vent Noble Gas M R(3) g(2)

a. D Activity Monitor Iodine Sampler W NA NA NA b.

NA NA

c. Particulate W NA Sampler

d. System Effluent D NA R Q Flow Rate Device

e. Sampler Monitor D HA R Q Flow Rate Measurement

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29. New Specification: (Cont.) Page 99 Table 4.20-1 (Continued)

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

1. Instrument indicates measured levels above the alarm / trip setpoint. 1 2.. Circuit failure.

3. Instrument indicates a downscale failure. ,

4. Instrument controls not set in operate mode-.

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

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

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument\ controls not set in operate mode.

l (3) The INSTRUMENT CHANNEL CALIBRATION shall be performed using one or more reference standards.

(4) A check shall be perfonned prior to each release.

(5) A check shall be performed prior to each release via a Waste Gas Decay Tank (s).

(6) To be performed when device is accessible and conditions do not pose a personnel safety hazard (i.e., potential main steam safety actuation).

(7) The CHANNEL TEST shall also demonstrate that thc Waste Gas System automatically isolates and that control room annunciation occurs if any of the following conditions exist:

1. Instrument indictes measured levels above the alann/ trip setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

Discussion:

The changes to Table 4.20-1 are made to be pursuant with the Standard Radiological Environmental Technical Specifications, t - - ____________

FACII.1TY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 100

30. Existing Specification:

4.21 LIQUID EFFLUENTS 4.21.1 Concentration Surveillance Requirements The concentration of radioactive material at any time in liquid effluents released from the site shall be continuously monitored in accordance with Table 3.15-1.

The liquid effluent continuous monitor having provisions for automatic termination of liquid releases, as listed in Table 3.15-1, shall be used to limit the concentration of radioactive material released at any time from the site to areas beyond the site boundary to the values given in Specification 3.17.3.

The radioactivity content of each batch of radioactive liquid waste to be discharged shall be determined prior to release by sampling and analysis in accordance with Table 4.21-1. The rest 0.ts of pre-release analyses shall be used with the calculational methods in the ODCM t6sassure that the concentration at the point of release is limited to the values in Specification 3.17.1.

Post-release analyses of samples from batch releases shall be performed in accordance with Table 4.21-1. The results of the post-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release are limited to the values in Specification 3.17.1.

Bases This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to areas beyond the site boundary will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, 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 Objectives of Appendix 1,10 CFR Part 50, 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 radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water ;

using methods described in International Commission on Radiological I Protection (ICRP) Publication 2. I l

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5 FACILITY CHANGE SAFEIT ANALYSIS LOG NO. 921 PROPOSED AMB(DMENT No.155 PAGE 101

30. Existing Specification: (Cont.)

There are no continuous releases of radioactive material in liquid effluents from the plant. All releases frca the plant are by batch method.

New Specification:

4.21' LIQUID EFFLUENTS 4.21.1 Concentration Surveillance Requirements The concentration of radioactive material at any time in liquid effluents released from the site shall be continuously monitored'in accordance with Table 3.25-1.

The liquid effluent continuous monitor having provisions for automatic termination of liquid releases, as 1$sted in Table 3.15-1, shall be used to limit the' concentration of radioactive material released at any time from the site to areas beyond the site boundary to the limits given in Specification 3.17.1.

.: The radioactivity content of each batch of liquid effluent to be b discharged shall be determined prior to release by sampling and analysis in accordance with Table 4.21-1. The results of pre-release analyses shall be used with the calculational methods in the OFFSITE DOSE CALCULATION MANUAL (ODCM) to assure that the concentration at the poir,t of. release is limited to the limits of Specification 3.17.1.

Bases This Specification is provided to ensure that the concentration of radioactive material released in liquid waste effluent from the site to areas beyond the site boundary for liquid effluent will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within the limits of 10 CFR Part 20.106 to MEMBER (S) 0F THE PUBLIC. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotopes and its MPC f.n air (submersion) was converted to any equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

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4 FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 102

30. New Specification (Cont.)

There are no continuous releases of radioactive material in liquid effluents from the plant. All radioactive liquid effluent releases from the plant are by batch method. _

Discussion:

The changes here represent compliance with 10 CFR 20 Appendix B, Table II, Column 2, which does not guarantee that a licensee will meet the numerical guides for design objectives of 10 CFR 50, Appendix I.

The surveillance requirements relate to the concentration LCO (3.17.1).

The bases for the LCO is in the Standard RETS which assumes that for a

" Standard PWR" compliance with the maximum permissible concentration (MPC) limit on an hour by hour basis will also result in the picnt operation being ALARA in terms of the numerical guides for the design objectives of 10 CFR 50, Appendix I. This assumption is incorrect for the site specific environmental setting of Rancho Seco, therefore reference to 10 CFR 50, Appendix I compliance is removed from the bases statement in surveillance standards.

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' 31.' Existing Specification: Page 103 i

Table 4.21-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Analysis Type of Activity Lower Limit Liquid Release

Frequency Frequency Analysis Of. Detectic Type (LLIJ)

(uCf/ml)(a:

Each Batch Each Batch' Mn-54, Fe-59, 5 x 10-7 A. Batch Wasta(Re-lease Tanks b) Co-58, Co-60 P P In-65, Mo-99, Cs-134, Cs-137 Ce-141, and Ce-144 (c)

I-131 1 x 10-6

\

One Batch /M M Dissolved and 1 x 10-5

.i Entrained Gases (Gamma Emittars)

M ~

Each Batch Composite (d) H-3 1 x 10-5 P

Gross Alpha 1 x 10-7 4 (d)

Each Batch Composite Sr-89, Sr-90 5 x 10-8 p .

W----____.__._ _ _ _ _ _ _ _ _ _

21. Existing Specification: (Cont.) Page 104 Table 4.21-1 (Continued)

RADIOACTIVE LIQUID WASTE SN4PLIhG AND ANALYSIS PROGRAM Table Notation

a. The lower limit of detection (LLD) is defined in the ODCM.

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

Prior to sampling, each batch will be isolated and thoroughly mixed, to assure representative sampling.

c. Other peaks which are measureable and identifiable, together with the

' listed nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analysis should not be reported as being present at the LLD level.

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

e

31. New Specification: Page 105-Table 4.21-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Analysis Type of Activity Lower Limit Liquid Release Frequency Frequency Analysis Of Detection Type (LLD')

(uCi/ml)(a)

Each Batch Mn-54, Fe-59,- -5 x 10-7 A. Batch Waste lease Tanks b.d(Re-) Each Batch Co-58, Co-60 P P Zn-65, Mo-99, Cs-134, Cs-137 Ce-141, and Ce-144 (c)-

I-131 3 x 10-7' Dissolved and 1 x 10-5

\

Entrained Gases

-(Gamma Emitters) il H-3 1 x 10-5

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31. New Specification: (Cont.) Paga 106 L

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RANCHO SECO UNIT 1 l l TECHNICAL SPECIFICATIONS l j

- Surveillance Standards l

TABLE 4.21-1 (Continued)

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM l l.

Table Notation ;

a. ;(1) The lower' limit of detection (LLD) for a radionuclides presented in l this table is the largest concentration, expressed in microcuries l per milliliter, which is required to be detected, if present, in l order to achieve compliance with the limits of Specification 3.17.1 i (10CFR20, Appendix B. Table II, Column 2). l (2) The LLD of a radioanalysis system is that value which will indicate the presence or absence of radioactivity in a sample when the probability of a false positive and of a false negative u determination is stated. .The probabilities of false positive and l false negative arsAtaken as equal at 0.05. The equation for LLD in - I microcuries per milliliter is given by the equation:

l LLD = f[2.7 + 3.29(Br)-0.5] l J./LHYLVij -

A where 2.7 = factor to correct for Poisson I distribution at very low background count rates. j l' !

f = correction factor' to account for systematic j errors = 1.1. j i

B = background (counts) :

i r=1+ts (ts 1 th) t b

tb = background count time (seconds) ts = sample count time (seconds)

'3.7E4 = disintegrations /second/ picocurie Y = yield of radiochemical process

~

E = counting efficiency (disintegrations / count)

/

Y = sample volume (liters) or mass (kilograms) :

I

31. New Specification:.(Cont.) Page 106a TABLE 4.21-1 (Continued).

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM T = [1-exp(-Atc)exp(-Atc)3 A

where A = decay. constant (seconds -1) tc = time from collection to start of counting

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

Prior to sampling, each batch will be isolated.. and then thoroughly mixed, to assure representative sampling.

c. Other peaks which are measureable and identifiable, together with the listed nuclides, shall also be identified and reported. Nuclides which are not observed for the analysis shall be reported as "less than" the 1 instrument's LLD, and shall not be reported as being present. The "less than" values shall not be used in the ODCM evaluations. However, if the nuclide is measured and identified at a value less than the Table 4.21-1 LLD value, it shall be reported and entered into the ODCM evaluations.

d. Miscellaneous Water Evaporator release is via the gaseous pathway.

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FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 107 1

31. New Specification: (Cont.)

Discussion .

Table notation items are clarified for wording (a., b.'and c.). A new set of LLD's are established at a concentration equivalent of about 50% of the values in Technical Specification 3.17.2, except for H3, which is at a concentration equivalent of less than 1%. (Refer to Attachment B for added detail on LLD changes). Appendix I compliance was removed due to the fact it relates to dose and not concentrations. Added conservatism was also included by increasing the sampling and minimum analysis of dissolved and entrained gases (gamma emitters). Clarification is made that the LLD's are for 10 CFR 20 compliance only and the requirement to perform composite analysis is deleted.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 '

PROPOSED AMENDMENT No. 155 PAGE 108

32. Existing Specification 4.21.2 Doses Dose Calculations ,

Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose calculation Manual (ODCM) at least monthly.

Bases This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and, at the same time, implement the guides set forth in Section IV,A of Appendix I to assure that the releases of Radioactive material in liquid effluents will be kept "as low as reasonably achievable." The Dose Calculations Methodology in the ODCM implements the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified 1 in the ODCM for calculating the doses due to the actual release rates of radioactive. materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and Regulatory Ouide 1.113, " Estimating aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

New Specification:

4.21.2 Dosen Dose Calculations Cumulative dose contributions and, cumulative dose projections associated with the release of liquid RADI0 ACTIVE EFFLUENTS from the site (see Figure 5.1.4) shall be determined in accordance with the sampling and analyses specified in Table 4.21-2 and the methodology described in the Offsite Dose Calculation Manual (ODCM) at the following frequencies:

a. Prior to the initiation of a release of liquid RADIOACTIVE EFFLUDTT; and,

I FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 l PROPOSED AMENDMENT No. 155 PAGE 109

32. New Specification (Cont.)

b. Weekly, based on gamma-emitter and tritium analyses of RADIOACTIVE EFFLUENTS released during the previous seven days; and,

c. Monthly, based on gamma-emitter and tritium analyses of RADIOACTIVE EFFLUENT releases during the previous calendar month and the results of analyses performed on composite samples shall be added to the monthly dose calculation.

A dose tracking system and administrative dose limits shall be established and maintained. Operating parameters shall be adjusted l in accordance with methodology described in the ODCM such that the dose values at any time, when projected to the end of the applicable time period, do not exceed the doses specified in Technical Specification 3.17.2.

Bases This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. Tha ;

Limiting Condition for Operation implements the guides set forth 13 Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and, at the same time, implement the j 5

guides set forth in Section IV.A of Appendix I which assures, by j definitica, that the releases of radioactive material in liquid j effluents will be kept "as low as reasonably achievable." The dose j calculations methodology in the ODCM implements the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculation procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. I The equations specified in the ODCM for calculating the doses due to !

the actual release rates of radioactive material in liquid effluents are consistent with the methodology provided in Regulatory Guide l; 1.109, " Calculation of Annual Dose to Man from Routine Releases of l Reactor Effluent for the Purpose of Evaluating Compliance with 10 CFR l Part 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.13, " Estimating Aquatic Dispersion of Effluents from Accidental and {

Routine Reactor Releases for the Purpose of Implementing Appendix I," I April 1977. l l

The results from composite samples during the period 1981 through j 1984 indicates that Cs-137, Cs-134, Co-58 and Co-60 constitute 80 i percent of the historical mix of gamma emitting radionuclides in j plant liquid effluents. Another 13 percent consists of I-131. When l the thyroid is separated as a limiting organ, 97.8 percent of the

FACILITY CHANGE SAFE 1T ANALYSIS LOG NO. 921 ,

PROPOSED AMENDMENT NO. 155 PAGE 110

32. New Specification (Cont.)

total body dose and 97.6 percent of the limiting organ dose are due to Cs-134 and Cs-137. Essentially 100 percent of the thyroid dose is due to I-131. j i

The activity analysis of Cs-134, Cs-137 and I-131 at the Lower Limits of Detection specified in Table 4.21-1 are based on an estimated annual plant radioactive effluent outflow of 20 million gallons per year with an average dilution flowrate of 5,000 gallons per minute.

These Lower Limits of Detection provfie an adequate basis for {

determining the presence or absence of dose due to other radionuclides in plant liquid effluents, when no other indications are revealed during sample analysis.

The dose tracking system ensures that the dose limits prescribed in Technical Specification 3.17.2 will not be exceeded at the 95 percent confidence level. The methodology presented in the ODCM provides for adjustment of operational and analysis parameters to factor in variables such as annual radiological liquid effluent relcase volume, discharge canal flow rate, and current cumulative dose.

The dose tracking system provides for prompt updating of cumulative dose and contains feedback mechanisms to assure that the target dose values are not exceeded. The tracking system also contains review ,

3 and approval of batch radiological liquid effluent releases at multiple management levels.

There is also reasonable assurance that the operation of the facility will not result in radionuclides concentrations in finished drinking water that are in excess of the requirement of 40 CFR 141.

Discussion:

The proposed amendment addresses the NRC concern in the NRC Inspection Report 86-15 which states the Rancho Seco LLD valves are in error since the use of these values can result in releases of radioactive materials to which offsite doses may be attributed. These doses may be in excess of the limits provided to implement 10 CFR 50, Appendix I, and 40 CFR 190.

The details of the dose tracking surveillance and new dose methodology is described in the ODCM. The addition of the LLD values in Table 4.21-2 are specified in Attachment B of this safety evaluation. Historical data was used in determining the LLD values and indicated that Cs-134, Cs-137, I-131 and tritium could be used as indicators for all nuclides in the Rancho Seco effluent mix.

i

32. New Specification: (Cont.) -

Page 111 1

Table 4.21-2 l RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Analysis Type of Activity Lower Limit -

Liquid Release Frequency Frequency Analysis Of Detection Type (LLD)

(uCi/ml)(a)

Each Batch Each Batch Cs-134 4 x 10-8 A. Batch Waste (

lease Tanks Cs-137 5 x 10-8 P P I-131 1 x 10-7 H-3 1 x 10-5 Each Batch Composite (d) H3 1 x 10-5 P M Na24 3 x 10-6 Cr51 3 x 10-5 Mn54 1 x 10-7 7

Fe59 CoS7 4 x 10 6 CoS8 4 x 10 6 Co60 11 xx 10 10-7 Zn65 5 x 10-7 Sr89 3 x 10-8

\ Sr90 3 x 10-8 Zr95 6 x 10-7 g

Nb95 1 x 10-8 Mo99 4 x 10-5 Ag110m 1 x 10-7 1131 1 x 10-8 1133 5 x 10-6 Cs134 9 x 10-9 Cs136 4 x 10-8 Cs137 7 x 10-8 Ba140 4 x 10-8 La140 6 x 10-6 Cel41 5 x 10-7 Cel44 5 x 10-7 Gross Alpha 1 x 10-7

32. New Specification: (Cont.) Page 112 I

i RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS l

, Surveillance Standards TABLE 4.21-2 (Continued) -

l RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM J l

Table Notation j The lower limit of detection (LLD) for a radionuclides presented in I

a. (1) this table is the largest concentration, expressed in microcuries l per milliliter, which is required to be detected, if present, in !

order to achieve compliance with the limits of Specification 3.17.2 i l (10CFR50, Appendix I). : !

(2) The LLD of a radioanalysis system is that value which will indicate I the presence or absence of radioactivity in a sample when the Ii probability of a false positive and of a false negative j i determination is stated'. The probabilities of false positive and l false negative are taken as equal at 0.05. The equation for LLD in j microcuries per m*111111ter is given by the equation: i ,

i LLD = f[2.'7 + 3.29(Br)-0.5) l l

3.7E4(YEVT) - -

where 247 = factor to correct for Poisson distribution at very low background count rates. ,

g i f = correction factor to account for systematic errors = 1.1. ;

B = background (counts) l! I r=1+ s (ts<tI b t

b l tb = background count time (seconds) 3 ts = sample count time (seconds) ; ,

3.7E4 = disintegrations /second/ picocurie l Y = yield of radiochemical process i

/ E = counting efficiency (disintegrations / count)

V = sample volume (liters) or mass (kilograms)

32. New Specification

Page 112a TABLE 4.21-2 (Continued)

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation T = [1-exp(-itelexp(-itr)]

k where A = decay constant (seconds -1)'

tc = time from collection to start of counting

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

Prior to sampling, each batch will be isolated, and then thoroughly mixed, to assure representative sampling. ,

c. Other peaks which are measureable and identifiable, together with the listed nuclides, shall also be identified and reported. Nuclides which are not observed for the analysis shall be reported as "less than" the instrument's LLD, and shall not be reported as being present. The "less than" values shall not be used in the ODCM evaluations. However, if the nuclide is measured and identified at a value less than the Table 4.21-1 l LLD value, it shall be reported and entered into the ODCM evaluations.

d A composite samp1'e is one in which the quantity of liquid samples is proportional to the quantity of liquid waste discharged and in which the

,g~

method of sampling employed results in a specimen which is representative

~

of the liquids released.

l

FACILITY CHANGE SAFETY ANALYSIS LOG NO. -921 PROPOSED AMENDMENT No. 155 PAGE 113

33. Existing Specification:

4.21.3 LIQUID HOLDUP TANKS

The quantity of radioactive material contained in each tank listed in Specification 3.17.3 shall be determined to be within the specified limit by analyzing a representative sample of the tank's contents at least weekly when radioactive materials are being added to the tank.

Bases Restricting the quantity of radioactive material contained in the specified outdoor tans provides assurance that in the event of an uncontrolled release of the contents, the concentration at the nearest surface water supply in an unrestricted area would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2,

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

Bi New Specification:

4.21.3 LIQUID HOLDUP TANKS

Surveillance Requirements The quantity of radioactive material contained in each tank listed in Specification 3.17.3 shall be determined to be within the specified limit by analyzing a representative sample of the tank's contents at least weekly when radioactive materials are being added to the tank.

Bases Restricting the quantity of radioactive material contained in the specified outdoor tanks provides assurance that in the event of an uncontrolled release of the tank's contents, the concentration at the nearest potable water supply and the surface water supply and the surface water supply in an unrestricted area would be less than the limits of 10 CFR Part 20, Appendix B. Table II, Column 2.

Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the

4 FACILITY CHANGE SAFETT ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 114

33. New Specification: (Cont.)

tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system or the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM.

Discussion Addition is made to the bases that makes assurance that potable water is ;

to be within the 10 CFR 20 Appendix B' concentration limits for radioactive. !

materials in.the event of an uncontrolled liquid holdup tank release.

\

p-i 1

l l

I l

FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 115

34. Existing Specification:

N/A New Specification:

4.21.4* LIQUID EFFLUENT RADWASTE TREATMENT Surveillance Requirements Doses due to liquid releases to unrestricted areas shall be projected at least once per 31 daye in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (0DCM) when LIQUID

. EFFLUENT RADWASTE TREATMENT SYSTEMS are not being fully utilized.

The' installed LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM shall be consideted OPERABLE by meeting Specification 3.17.1 and 3.17.2.

Bases The OPERABILITY of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM ensures that this system will be available for use whenever liquid effluents taquire treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when y specified provides assurance that the release of radioactive '

materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Derign Criterion 60 of Appendix A to 10 CFR Part 50, and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid efficents.

The installation of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM is not complete. This specification will become effective when the system is declared operable.

Discussion l The new surveillance requirement is added per Dfstrict commitment to the NRC, assuring the operability of the radwaste treatment system in the event of liquid effluent release.

l L_______________ _ _ _

35. Existing Specification: Page 116 4.22 GASEQUS EFFLUENTS 4.22.1 Dose Rate Surveillance Requirements The release rate of noble gases in gaseous effluents shall be controlled by the offsite dose rate as established in Specification 3.18.1. The dose rate -

shall be detemined in accordance with the ODCM.

The noble gas effluent continous monitors, as listed in Table 3.16-1, shall use monitor setpoints to limit offsite doses within the values established in Specification 3.18.1. .

The release rate of radioactive materials, other than noble gases, in gaseous effluents shall be detemined by obtaining representative samples and perfoming analyses in accordance with the sampling and analysis program, specified in Table 4.22-1.

The dose rate at and beyond the site boundary, due to Iodine-131, tritium, and all radionuclides in particulate fonn with half-lives greater than 8 days released in gaseous effluents, .shall be determined to be within the required f limits by using the results of the sampling and analysis program, specified in

't Table 4.22-1, in perfoming the calculations of dose rate beyond the site boundary in accordance with the ODCH.

Bases -

This specification is provided to ensure that the dose rate at any time at the site boundary from gaseous effluents will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material ,

discharged in gaseous effluents will not result in the exposure of an individual outside the restricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits spe:ified in Appendix B 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 in the atmospheric diffusion factor above that for the restricted area boundary. The specified release rate limits restrict, at all times, the correspondir" gamma and beta dose rates above background to an individual at or beyond the restricted area boundary to 500 mrem / year to the total body or to 3,000 mrem / year to the skin. These " release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1,500 mrem / year.

s

{ J

35. New Specification: Page 117 4.22 GASEOUS EFFLUENTS i 4.22.1 Dose Rate Surveillance Requirements l

The dose rate due to noble gases in gaseous effluents shall be detemined to be within the limits in Specification 3.18.1 in accordance with the methodology described in the 0FFSITE DOSE CALCULATION MANUAL (00CM).

The noble gas effluent continuous monitors, as listed in Table 3.16-1, shall use monitor setpoints to limit the dose rate in unrestricted areas to the limits in Specification 3.18.1.

The release rate of radioactive 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 4.22-1.

The dose rate due to Iodine-131, Iodine-133, tritium, and all radioactive material in particulate form with half-lives greater than 8 days released in gaseous effluents, shall be determined to be within the limits in Specification 3.18.1 by using the results of the sampling and analysis program specified in Table 4.22-1, and in accord,ance with the methodology described in the ODCM.

Bases i

This specification is provided to ensure that the dose rate at any. time at the Exclusion Area Boundary (Figure 5.1-1) from gaseous effluents will be within ;

the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)(1)). For individuals who may at times be within the Exclusion Area Boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the Exclusion Area Boundary to less than or equal to 500 mrem / year to the total body or to less than or equal to 3,000 mren/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the grass-cow-milk-infant pathway to less than er equal to 1,500 mrem / year for the nearest dairy cow to the plant.

.36. Existing Specification: . Page 118 Table 4.22-1 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Analysis Type of Activity Lower Limit ,

Gaseous Release Frequency Frequency Analysis of Detection i (LLD) a j Type _

(uCi/mi)

A. Waste Gas Each Tank Each Tank Principal Gamma 1 x 10-4 ;

(f) i Storage Tank Grab P Emitters Sample P

B. Containment Each Purge Each Purge Principal Gamma 1 x 10-4 Purge Grab Emitters (f)

Sample (e) P P H-3 1 x 10-6 C. Auxiliary Mlb,c) M(tr) Principal Gamma 1 x 10-4 N Building Stack, Grab Emitters - (f) /

and Radwaste Sample ,

Service Area H-3 1 x 10 5 Yent

- D. All Release Continuous W(d) I-131 1 x 10-12 Types as listed Charcoal in A,B,C above Sample Continuous Wlo)

Particulate Principal Gamma 1 x 10-11 Sample Emitters (f)

(I-131, Others)

- Continuous M Gross Alpna 1 x lu-Al Composite Particulate Sample I

Continuous Q Sr-89, Sr-90 1 x 10-11 Composite Particulate ;

Sample Continuous Noble Gas Noble Gases 1 x 10-4 Monitor Beta or Gamma as Xe-133 (Gros s)

- : . g . .

's. -

36. Existing Specification: (Cont.) Page 119 Table 4.22-1 (Continued)

RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation,

~

a. The lower limit of detection (LLD) is defined in the 00CM.

, b. Analysis shall also be perforced when gross beta-gamma activity analysis of reactor coolant indicates greater than 10 uCi/ml and after each 10 uCi/ml increase in the gross beta-gamma activity analysis. :

c. Tritium grab samples shall be taken at least once per seven days from the ventilation exhaust from the auxiliary building stack during refueling and anytime fuel is in the spent fuel pool and the pool temperature exceeds 110*F. Below 110*F there is essentially no evaporation from this sou rce.

d. Samples shall be changed at least weekly with analyses to be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Sampling and analysis shall also be performed when reactor coolant indicates 10gCi/ml gross beta gamma activity and every 10nci/ml increases thereaf ter. When samples collected for less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the. corresponding LLD's may be increased by a' factor

~

k.. - of 10. '

n e. Tritium grab samples shall be taken at least daily during refueling activities.

f. Principle gamma emitters for which the LLD applies are: Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-135m for gaseous samples and Mn-54, Fe-59, Co-58, Co-60, Mo-99 (or Tc99m), Cs-134, Cs-137, Ce-141, and Ce-144 for particulate samples. This list does not mean only these nuclides will be reported; other peaks that are measurable and identifiable will alsc be

< reported. i 1

y s- :.

36. Nrv Sp*cification:

' ~ ~ ~ ~

Page 120 r.% . A is, ;. & EGu 4J.idIU$d RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.22-1 RADI0 ACTIVE GASEOUS WASTE sal @ LING AND ANALYSIS PROGRAM Sampling Minimum Analysis Type of Acttvity Lower Limit Gaseous Release Frequency Frequency Analysis of Detection Type (LLD) a (uCi/ml)

A. Waste Gas P P Storage Tank Each Tank Each Tank Principal lama 1 x 10-4 Grab Emitters (f)

Sample B. Reactor Building P P Purge Vent Each Purge Each Purge (b,e,1) Principal Gama 1 x 10-4 Grab. Emitters (f)

Sample (b,e,1 )

H-3 1 x 10-6 C. Auxiliary M(Q.,c.e) M(b) Principal Gama 1 x 10-4 Building Stack - Grab Emitters (f)

B- H-3 1 x 10-6 D. Auxiliary M('b) M(b) Principal Gama 1 x 10-4 Building Grade Grab Emitters (f)

Level Vent Sample H-3 1 x 10-6 E. All Release Continuous W(d) 1-131 1 x 10-12 Types as listed Charcoal j in A,B,C,0 above Sample I-133 1 x 10-10 Continuous W(d)

Particulate Principal Gama 1 x 10-11 Sample Emitters (f)

(I-131, Others) ..

Continuous N Gross Alpha (h) 1 x 10-AA Composite Particulate Sample Sr-89, Sr-90(g) 1 x 10-11 Continuous Noble Gas Noble Gases 1 x 10-6 Monitor Gross Beta and Gamma as Xe-133

h : 36. - Nw Specification: . (Cont. ) Pega' 121 c

' RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS -

Surveillance Standards ,

l Table 4.22-1 (Continued)

i RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM C 1

Table Notation

a. (1) The lower limit of detection (LLD) for a radionuclides presented in this table is the largest concentration, expressed in microcuries per milliliter, which is required to be detected, if present, in order to achieve compliance with the limits of Specifications 3.18.1, 3.18.2 and 3.18.3.

(2) The LLD of a radioanalysis system is' that value which'will indicate the presence or absence of radioactivity in a sample when the probability of a false positive and of a false negative determination is stated. The probabilities of false positive and [

false negative are taken as equal at 0.05. The equation for LLO in microcuries per millil' iter is given by the equation:

gLLD = f[2.7 + 3.29(Br) 0.53 -

3./E4 HEVT).

L where 2.7 = factor to correct for Poisson distribution at very low background count rates.

f = correction factor to account for systematic errors = 1.1.

B = background (counts) t r=1+ s (ts 5 tb) '

t b

tb = background count time (seconds) ts = sample count time (seconds) 3.7E4 = disintegrations /second/ picocurie I

Y = yield of radiochemical process E = counting efficiency (disintegrations / count)

V = sample volume (liters) or mass (kilograms)

T=[1-exp(-Ats)exp(-itc)3 A

where 1. decay constant (seconds -1)

~ tc = time from collection to start of counting

36. New Specification: Pagn 121a l

RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS

\\

Surveillance Standards l i

Table 4.22-1

.g (Continued)

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation

b. Analysis shall also be performed Ehen gross beta or gamma activity anaylsis of reactor coolant indicates greater than 10 pCi/ml and after each 10 pCi/mi increase in the gross beta or gamma activity analysis.

c. Tritium grab samples shall be taken at least once per seven days from the ventilation exhaust from the auxiliary building stack during refueling !.

and anytime fuel is in the spent fuel pool and the pool temperature exceeds 110*F. Below 110*F there is essentially no evaporation from this '

source.

d. Samples shall be changed at least weekly and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Sampling and analysis shall also be performed when reactor coolant indicates 10pci/ml gross beta gamma activity and every 10pci/m1 increases theteafter. When samples collected for less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs maybe increased by a factor of 10.

. l

e.

Tritium grab samples shall be taken at least daily during refueling activities.

f. Principle gamma emitters for which the LLD applies are: Kr-87, Kr-88, l Xe-133, Xe-133m, Xe-135, Xe-135m, and Xe-138 for gaseous samples and i; Mn-54, Fe-59, Co-58, Co-60, In-65, Mo-99, (or Tc99m), Cs-134, Cs-137, l' Ce-141, and Ce-144 for particulate samples. This list does not mean only these nuclides will be detected and reported. Other peaks that are ,

measurable and identifiable shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.2.3. Nuclides which are below the LLD for the analysis shall be reported as "less than" the nuclide's LLO and shall not be reported as a being r/esent at the LLD level for that nuclide. However, if the nuclide is measured and identified at a value less than its predetermined LLD !

t value, it shall be reported and entered into the ODCM evaluations.

g. Gross beta analysis performed on a monthly basis for each environmental 'l release particulate sample. If any one of these samples indicates i greater than 1.0 E-11 pCi/cc gross beta activity then a Sr-89, Sr-90 ;

analysis will be performed on those samples exceeding this value. ,

h. Gross beta performed on a monthly basis for each environmental release particulate sample. This fulfills the requirements of performing a monthly composite.

i. After purging seven reactor building volumes, sampling and analysis of Reactor Building Purge Yent exhaust shall be conducted at least once per seven days.

FACILITY CHANGE SAFEIT ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 122

37. Existing Specification:

4.22.2 NOBLE GASES Dose Calculations Cumulative air dose contributions for the quarterly or yearly period as applicable shall be determined in accordance with tl e Offsite Dose Calculational Manual (ODCM) at least monthly.

Bases This specification is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as reasonably achievable." The Surveillance Requirements 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 s on models and data such that the actual exposure of.

an individual through the appropriate pathways is unlikuly to be substantially underestimated. The dose calculations established in 1 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 for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releascs from Light-Kater-Cooled Reactors," Revision 1 July 1977. The ODCM equations provided for determining thec air dosee at the site boundary will be based upon the historical average atmospheric conditions. NUREG-0133 provides methods for dove calculations consistent with Regulatory Guides 1.109 and 1.111.

New Specification:

4.22.2 DOSE-NOBLE GASES Dose Calculations Cumulative air dose contributions for the calendar quarter and l

calendar year shall be determined in accordance with the methodology described in the OFFSITE DOSE CALCULATIONAL MANUAL (ODCM) at least monthly.

+

FACILITY CHANGE SAFE 1T ANALYSIS LOG NO. 921 i PROPOSED AMENDMENT NO. 155 PAGE 123

37. New Specification: (Cont.)

Bases This specification is provided to implement the requirements of Sections II.B III. A, and IV.A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as reasonably achievable." The Surveillance Requirements laplement the requirements in Section III.A of Appendix I that i conformance with the guides of Appendix I be shown by calculational i procedures based on models and data such that the actual exposure of an individual throuFh 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 are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,

" Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at the Site Boundary For Gaseous Effluents (Figure 5.1-3) and are based upon the historical average atmospheric conditions.

Discussion The changes here represent conformance with the Standard RETS.

Clarification is made for the frequency of determination of the cumulative air dose contributions.

38. Existing Specification: Page 124 4.22.3 Iodine-131, Tritium and Radionuclides in Particulate Fom Dose Calculations Cumulative dose contributions for the quarterly or yearly period as applicable shall be detemined in accordance with the 00CM at least monthly.

Bases .

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. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactiv'e materials in gaseous effluents will be kept "as low as reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods approved by the NRC for calculating the doses due to the actual release rates of the subject f, materials are required to be consigtent with the methodology provided in b Regulatory Guide 1.109, '" Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 s CFR Part 50, Appendix I," Revision 1, October 1977 ana Kegulatory Guide 1.111

" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1 July 1977. These equations also provide for detemining the actual doses bdsed upon the historical average atmospheric conditions. The release rate specifications for radioiodines, radioactive material in particulate fem and radionuclides other than noble gases are dependent on the existing radionuclides pathways to man, beyond the site boundary. The pathways which i are examined in the development of these calculations are: (1) indivicual I 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. i l

l I

l l

l

38. New Specification: Page 125 4.22.3 Dose-Iodine-131, Iodine-133, Tritium, and Radioactive Materials in 4 l

Particulate TTrm_.

l Dose Calculations Cumulative dose contributions for the calendar quarter and calendar year period shall be determined in accordance with the methodology described in the (0DCM) 0FFSITE DOSE CALCULATION MANUAL at least monthly.

I Bases 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. The ACTION l statements provide the required operating flexibility and at the same time I

implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable." The ODCM calculational methods specified in l the surveillance requirements implement the requirements in Section III.A of i

Appendix I that conformance with the guides of Appendix I be shown by

! calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the inethodology provided in Regulatory Guide 1.109,

" Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents 1

for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appcndix I,"

Revision 1, 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 1977. These equations also provide for estimating doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, and radioactive material in particulate form are dependent on the existing radionuclides pathways to man at or beyond the Site Boundary for Gaseous Effluents (Figure 5.1-3). 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.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 126

39. Existing Specification:

4.23 GASEOUS RADWASTE TREATMENT Dose Projections Doses due to gaseous releases beyond the site boundary shall be projected at least monthly in accordance with the ODCM.

Bases The operability of the gaseous radwaste treatment system and the ventilation exhaust treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable."

This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and Design Objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were spe'cified as a suitable fraction of the guide cet forth in Sections II.B and II.C of Appendix I, 10 CrR Part 50, for gaseous effluents.

New Specification:

4.22.4 GASEOUS RADWASTE TREATMENT Surveillance Requirement Doses due to gaseous releases to areas at and beyond the Site Boundary For Gaseous Effluents (see Figure 5.1-3) shall be projected at least once per 31 days in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM) when Gaseous Radwaste Treatment Systems are not being fully utilized.

The installed VENTILATION EXHAUST TIEATMENT SYSTEM and Waste Gas System shall be considered OPERABLE by meeting Specifications 3.18.1, 3.18.2 and 3.18.3.

Bases The operability of the Waste Gas System and the VENTILATION EXHAUST TREATMENT SYSTEMS ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of l 1

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' FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED A)ENDMENT NO. 155 PAGE 127

39. New Specification: (Cont.)

systems be used, when specified, provides reasonable assurance that the releases of radioretive materials in gaseous effluents will be kept."as low as reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of.the-systems were specified as the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

Discussion These changes are pursuant with the Standard RETS. Additional clarification is made for surveillance standards.

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FACILITY CHANGE SAFETY ANALYSIS' LOG NO. 921 PROPOSED AMENDMENT NO. 155- ,

PAGE 128

40. Existing Specification:

-4.24 GAS STORAGE TANKS Surveillance Requirements The quantity of radioactive material contained in~each gas storage-tank shall be determined to be within the limit of 3.20 at least daily when radioactive materials are being added to the tank and the Reactor Coolant System activity exceeds the limits of Specification 3.1.4.

Bases Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest site boundary will not exceed 500 mrem.

This is consistent with Standard Review Plant 15.7.1, " Waste Gas System Failure."

New Specification:

\

4.22.5 GAS STORAGE TANKS Surveillance Requirements The quantity of radioactive material contained in each waste gas decay tank shall be determined to be within the limit in Specification 3.18.5 at least daily when radioactive materials are being added to the tank and the Reactor Coolant System activity exceeds the limits of Specification 3.1.4.

Bases Restricting the quantity of radioactivity contained in each waste gas decay tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the exclusion area boundary (see Figure 5.1-1) will not exceed 500 mrem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure."

Calculations have shown that the reactor coolant activity must exceed the limits of Specification 3.1.4 before the waste gas decay tank activity approaches the limits of Specification 3.18.5.

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FACILIIT CHANGE SAFETY ANALYSIS, -

LOG.NO. 921 l '- PROPOSED AIGNDMENT No.155 PAGE 129-l

40. New Specification: (Cont.)-

Discussion Clarification is made that the exclusion area boundary is the limiting boundary for determining dose to an individual from an uncontrolled release from the gas storage tanks.

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41. Existing Specification: Page 130 4.25 SOLID RADI0 ACTIVE WASTES Surveillance Requirements The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at lease one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM. .

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test" specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, ,as provided in Specification 6.15, to assure SOLIDIFICATION.of subsequent batches of waste.

Reports The semiannual Radioactive Effluent Release Report shall include the following information for each type of solid waste shipped offsite during the report period:

a. Container volume,

b. Total curie quantity (determined by measurement or estimate),

c. Principal radionuclides (determined by measurement or estimate),

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

e. Type of container (e.g., LSA, Type A, Type 8, Large Quantity), and

f. Solidification agent (e.g., cement, urea formaldehyde).

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4-' 41. Existing Specification: (Cont.)' Page 131

I 4.25 (Continued)

Bases This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion. 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not. limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times, aw-(, \

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41. New Specification: Page 132 N.

i 4.25- SOLID RADIOACTIVE WASTES

]

Surveillance Requirements

.1 4.25 T The solid radwaste systems shall be demonstrated OPERABLE at least j

s. once per 92 days by i

a. Operating the solid radwaste system at least once in the previous 92 days in.accordance with the PROCESS CONTROL PROGRAM, or b.- Verification' of the existence of a valid contract' for SOLIDIFICATION to be performed by a contractor in accordance with 'a PROCESS CONTROL PROGRAM.

4.25.2 The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at lease one representative test specimen from at least every tenth batch of each type.of wet radioactive waste (e.g.,

filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

-a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until suc.h time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent' test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from'each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.15, to assure SOLIDIFICATION of subsequent batches of waste.

Bases The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite.

This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

L - - _ _ _ _ _ ___i

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41. New Specification: (Cont.) Page 133 4.25 (Continued)

Bases The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid.radwastes require processing and packaging prior to being shipped offsite.

Discussion:

Changes here represent the addition of periodic ~ operability demonstrations for the solid radwaste system as recommended in the Standard RETS.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 134

42. Existing Specification:

4.26 RADIOLOGICAL ENVIRONMENTAL MONITORING Surveillance Requiremen_ts The radiological environmental monitoring samples shall be collected per Table 3.22-1 from the locations shown in the ODCM and shall be analyzed to the requirements of Tables 3.22-1 and 4.26-1.

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

The specified monitoring program is in effect at the present time.

Program changes may be initiated based on operational experience and changes in regional population or agricultural practices. The sample locations have been listed in the ODCM to retain flexibility for making changes as needed.

New Specification:

4.26 RADIOLOGICAL ENVIRONMENTAL MONITORING Surveillance Requirements The radiological environmental monitoring samples shall be collected per Table 3.22-1 from the locations shown in the RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) MANUAL and shall be analyzed to the requirements of Tables 3.22-1 and 4.26-1.

Bases The Radiological Environmental 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 implementsSection IV.B.2 of Appendix I to 10 CFR 50 and supplements the radiological effluent monitoring prcgram by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and ODCM modeling of the environmental

FACILITY GIANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 135

42. New Specification (Cont.)

exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The specified monitoring program is in effect at the present time. Program changes may be initiated based on operational experience and changes in regional population or agricultural practices. The sample locations have been listed in the REMP manual to retain flexibility for making changes as needed. ,

The detection capabilities required by Table 4.26-1 are .

state-of-the-art for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141.

Discussion:

Changes here represent that the REMP manual will be the governing document for the radiological environmental surveillance requirements. Reference to 40 CFR 141 is pursuant to the Standard RETS.

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43. New Specification: Page 137 AANCHO $ECO UNIT 1 TECHNICAL $ SPECIFICATIONS

$ surveillance Standards Table 4.26-1 MAXIMUM YALUES FOR THE LOWER LIMIT $ OF DETECTION (LLD)a, d Fish Milk food Products Mud and $11t Water Afrborne Particulate (pC1/kg, wet) (pC1/kg, wet) or Gases (pCf/m*) (pC1/kg, wet) (pC1/1)

Analyst s (pC1/1) '

gross beta 4(b). 1 x 10 2 3H 2000(1000(b)) ,

54 Mn 130 15 260 59re 30 150 130 58Co 15' 150 130 60Co 15 260 652n 30 952 r-Hb 15(t) 1 60 1311 1(b) 7 a 10-2 15 60 150 1 a 10-2(c) 130 134Cs 15 (10(b))

10 00 100 1 x 10-2(c) 150 137C s la (10(b))

15(el la0 Ba-La 15 (e)

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Discussion:

i Changes to Table 4.26-1 represent District conformance to the Standard RETS.

Lower maximum LLD values are defined for Cesium detection in drinking water and milk. In addition, LLD values are established for the liquid effluent pathway in mud and silt.for Cesium and cobalt detection.

The' addition of establishing LLD values for MUD and silt addresses the NRC concern in the July 22, 1986 staff evaluation that the mud and silt effluent pathway model did not take into account long-term buildup of concentrations of radionuclides in bottom sediments, thereby imparting doses to ingesting aquatic foods (bottom-feeding fish).

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 138

44. Existing Specification:

Table 4.26-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) a, d Table Notation

a. The LLD is defined in the ODCM.

Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLD's unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report,

b. LLD for drinking water.

c. LLD shown is for composite analysis. For individual samples, 5x10-2p ci/m3 is the LLD.

d. Other peaks which are measurable and identifiable, together with the nuclides in Table 4.26-1, shall be identified and reported.

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44. New Specii'..;ation: Pege 139 l

RANCHO SECO. UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards

(' '

Table 4.26-1 (Continued)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a, d Table Notation

a. (1) The lower limit of dhtection (LLD) for a radionuclides presented in this table is the largest concentration, expressed in microcuries per milliliter, which is required to be detected, i' if present, in order to achieve compliance with the applicable regulation, given stated operating conditions and calculation methology.

(2) The LLD of a radioanalysis system is that value which will indicate the presence or absence of radioactivity in a sample when the probab,ility of a false positive and of a false negatiye determination is stated. The probabilities of false positive and false' negative are taken as equal at 0.05. The l

equation for LLD in microcuries per mil.liliter is given by the -

s ,

equation: -

s LLO = f[2.7 + 3.29(Br)-0.5]

u.udtut.yiJ 1

where 2.7 = factor to correct for poisson distribution at verj, low background count rates.

f = correction factor to account for systematic errors = 1.1.

B = background (counts)

r=1+ t t

s (ts<t)b b

t b = background count time (seconds) t 3 = sample count time (seconds) 0.037 = disintegrations /second/ picocurie Y = yield of radiochemical process

[ E = counting efficiency (disintegrations / count) t V = sample volume (liters) or mass (kilograms)

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44. New Specification: (Cont.) 'Page 139a I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Table 4.26-1 (Continued)

)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a, d l Table Notation :

l l

T = [1-exp(-ite)exp(-itc))

A where A = decay constant (seconds -1) te = time from collection to start of counting (3) Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interferin'g nuclides, or other uncontrollable circumstances may render these LLD's unachievable. In such cases, the contributing J

factors will be identified and described in the Annual Radiological Environmental Operating Report. -

b. LLD for drinking water.

c. LLO shown is for composite analysis. For individual samples, 5x10-2pCi/m3 is the LLO.

1

d. Other peaks which are measurable and identifiable, together with the nuclides in Table 4.26-1, shall be identified and reported.

e. Total for parent and daughter.

Discussion:

Changes to Table 4.26-1 represent District conformance to the Standard RETS.

Lower maximum LLD values are defined for Cesium detection in drinking water and milk. In addition, LLD values are established for the liquid effluent pathway in mud and silt for Cesium and cobalt detection.

The addition of establishing LLD values for mud and silt addressed the NRC concern in the July 22, 1986 staff evaluation that the mud and silt effluent pathway model did not take into account long-term buildup of ;

concentrations of radionuclides in bottom sediments, thereby imparting doses to ingesting aquatic foods (bottom-feeding fish).

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 ;

i PROPOSED AMENDMENT NO. 155 PAGE 140 l l

l

45. Existing Specification: 4 l

4.27 IAND USE CENSUS Surveillance Requirements The land use census shall be conducted annually by using methods that will provide the best results, such as door-to-door survey, aerial l survey, or by consulting local agriculture authorities. l l

Reports i

The results of the land use census shall be included in the Annual l Radiological Environmental Operating Report. j Bases q i

This specification is provided to ensure that changes in the use of j areas at and beyond the SITE BOUNDARY are identified and that I 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 I that significant exposure

\ pathways via leafy vegetables will be identified and monitored, since a garden of this aize is the minimum a

required to produce the quantity (26 kg/ year) of leafy vegetable assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used: (1) that 20% of the garden was used for growing broad-leaf

. vegetation (i.e., similar to lettuce and cabbage), and (2) a j vegetation yield of 2 kg/ square meter. j I

New Specification: J 4.27 IAND USE CENSUS j i

I Surveillance Requirements The land use census shall be conducted annually by using methods that will provide the best results, such as door-to-door survey, aerial survey, or by consulting local agriculture authorities.

The land use census or portions thereof, shall be conducted during ,

the appropriate time of the year to provide the best results. )

l Reports j The results of the land use census shall be included in the Annual Radiological Environmental Operating Report. 3 1

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l FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 141

45. New Specification (Cont.)

Bases This specification is provided to' ensure that changes in the use of unrestricted areas are identified and that modifications to the RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL and ODCM are cade 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 vegetable assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used: (1) that 20% of the garden was used for growing broad-leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetarion yield of 2 kg/ square meter.

In addition, by gathering information on the liquid effluent pathway and the gaseous effluent pathway, the census will assure that proper radiological environmental monitoring and radioactive effluent controls are in ' place for adequate protection of the health and safety of the general public.

Discussion:

The changes represent the surveillance requirements as it relates to the LCO in Tech Spec 3.23 Land Use Census. Additions include liquid effluent pathway surveillance for current land use. Identification of the REMP as the environmental monitoring vehicle is included in the text of the Bases.

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46. Existing Specification: Page 142 l

. 4.29 FUEL CYCLE DOSE .

Surveillance Requirements 4

Cummulative dose contributions from liquid and gaseous effluents shall be detennined in accordance with Specifications 3.17.2.a,. 3.17.2.b, 3.18.1.a.

3.18.1.b, 3.18.2.a, 3.18.2.b, 3.18.3.a and 3.18.3.b, and in accordance with the Offsite Dose Calculation Manual (00CM).

Reports Special reports shall be submitted as required under Specification 3.25.

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

The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix 1. For the Rancho Seco site, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the plant remains within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of the annual dose to a member of the public to within the 40 CFRg 190 limits. For the purposes of the Special Report, it may be ass 0med that the dose commitment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is compl eted. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation which is part .

of the nuclear fuel cycle.

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46. New Specification: Page 143 4.29 FUEL CYCLE DOSE Surveillance Requirements Cumulative dose contributions from liquid and gaseous effluents shall be detemined in accordance with Specifications 4.21.2, 4.22.2, and 4.22.3 and in accordance with the OFFSITE DOSE CALCULATION MANUAL (00CM).

Cumulative dose contributions from direct radiation (including outside' storage tanks, etc.) shall be detemined in accordance with the OFFSITE DOSE CALCULATION MANUAL (00CM). This requirement is applicable only under conditions set forth in the Action Statement of Specification 3.25.

Reports .

Special reports shall be submitted as required under Specification 3.25.

Bases This specification is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the numerical guides for design objective doses of Appendix I or exceeds the reporting levels for the Radiological Environmental Monitoring Program. For the Rancho Seco site, it is unlikely that the resultant dose to a MEMBER OF THE PUBLIC .

! will exceed the dose limits of 40 CFR 190 if the plant remains within twice the numerical guides for design objectives of 10 CFR 50, Appendix I and if direct radiation (cutside storage tanks, etc.) is kept small. The Special Report will describe a course of action which should result in the limitation of the dose to a MEMBER OF THE PUBLIC for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any MEMBER OF THE PUBLIC is evaluated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation which is part of the uranium fuel cycle. -

Discussion

'Ihe modifications are in conformance with the Standard RETS'.

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. FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 144

47. Existing Specification:

5. DESIGN FEATURES 5.1 SITE Specification The Rancho Seco reactor is located on the 2,480 acres owned by Sacramento Municipal Utility District, 26 miles north-northeast of Stockton and 25 miles southeast of the City of Sacramento, California. FSAR figure 1.1-2 shows the plan of the site. The minimum distance to the boundary of the exclusion area, as defined in 10 CFR 100.3, shall be 2,100 feet. (1), (2)

REFERENCES (1) FSAR paragraph 1.2.1 (2) FSAR paraggaph 2.2.1 New Specification:

i

5. DESIGN FEATURES 5.1 SITE The Rancho Seco reactor is located on the 2,480 acres owned by Sacramento Municipal Utility District, 26 miles north-northeast of Stockton and 25 miles southeast of the City of Sacrament, California. The minimum distance to the boundary of the exclusion area, as defined in 10 CFR 100.3, shall be 2,100 feet.-

5.1.1 Exclusion Area The EXCLUSION AREA shall be shown in Figure 5.1-1.

5.1.2 Low Population Zone The LOW POPULATION ZONE shall ne shown in Figure 5.1-2.

5.1.3 Site Boundary for Gaseous Effluents The SITE BOUNDARY FOR GASEOUS EFFLUENTS shall be shown in Figure 5.1-3.

l FACILITY CHANGE SAFETY ANALYSIS- LOG No. 921.

PROPOSED AMIDIDtGNT NO.155 PAGE 145 4/.- New Specifict?.ien:';(Cont.)

5.1.4 Site Boundary for Liquid Effluents The' SITE BOUNDARY FOR LIQUID EFFLUENTS shall be shown in '

Figure 5.1-4.

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FACILITY G ANGE SAFI!T ANALYSIS LOG NO. 921 PROPOSED AMilNDMENT No. 155 .PAGE 150

47. New Specification (Cont.)

Discussion:

The addition of site schematic drawings identifying the affected areas of effluent release is pursuant with the Standard RETS.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 151

48. Administrative And Editorial Changes:

See Attachment A to this safety analysis which is a subset of the overall Technical Specification change of Proposed Amendment No.

155. The changes are made to the Table of Contents and Figures and Chapter 6, Administrative Controls.

Safety Analysis For Technical Specification Changes (Non-Administrative / Editorial) Items 1 - 47 The changes to the Technical Specifications will address the NRC staffs' review undertaken in connection with contamination found in the vicinity of the Rancho Seco plant. Technical Specification inconsistencies were found between the Lower Limit of Detection (LLD) as listed in Table 4.21-1 of the Technical Specification Surveillance Standards (Section 4.21.2) and the Technical Specification Sections (3.17.2 and 4.21.2) relating to 10 CFR Part 50, Appendix I design objectives. The NRC position stated that because of the highly atypical characteristics of the Rancho Seco cooling water system and of the receiving waters, liquid effluent releases with the current LLD valves, could result in excess of 10 CFR 50, Appendix I and limits specified in 40 CFR 190.

These concerns are a dressed in this safety analysis in the following groupings for adoption to Standard Radiological Effluent Technical ili Specifications (RETS) in NUREG-0472 and NUREG-0452 compliance with 10 CFR 50, Appendi: I requirements for liquid effluent releases:

1) Concentration Limits

2) Dose Limits

3) Radiological Environmental Monitoring Program

4) Land Use Census

5) Reporting, Procedures and Audits

6) Instrumentation

1. Concentration NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," provides calculational models for dose contributions for implementing 10 CFR Part 50, Appendix I "as low as is reasonably achievable" requirements. For liquid effluent releases, a near field average dilution factor is used which takes into account the maximum undiluted liquid waste flow, the combined liquid releases for each unit, and the mixing effects in the receiving water body in the near field of the discharge structure. For plants with non-recirculating main condenser cooling systems, the mixing effects in the receiving water body are ignored for conservatism. However, for plants with recirculating cooling systems, where cooling water discharge flow

FACILITY CHANGE SAFEIT ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 152

' Safety Analysis For Technical Specification Changes (Non-Administrative / Editorial (Cont.)

i

1. Concentration ]

rates are much less than for plants with non-recirculating cooling systems, credit is allowed for mixing effects in the near field of the receiving water body up to the degree of non-recirculating cooling water. j Rancho Seco has a recirculating main condenser cooling system. Based on a comparison on Environmental Statements for various nuclear power plants, the Rancho Seco design average discharge flow rate is one of the lowest of all U. S. nuclear power plants. As with similar plants, the liquid waste discharge includes condenser cooling and service water system blowdown, and other minor streams in addition to liquid radwaste effluents. However, atypically, at Rancho Seco there is little or no dilution of liquid wastes af ter discharge from the plant discharge structure due to the almost total absence of a receiving water body comprised of water other than from the plant discharge. Consequently, no credit is provided in the Rancho Seco ODCM for mixingg i n the receiving water body in the near field of the discharge structure.

3 The existing specification for concentration in the Rancho Seco Technical Specifications (RSTS) was based on the Standard Radiological Effluent Tech Specs (RETS) NUREG-0472, which assumed that for a " standard PWR" compliance with the Maximum Permissible

. Concentration (MPC) on an hour by hour release basis will result in the plant being ALARA. This assumption was incorrect due to the fact that Rancho Seco is a " dry" plant for effluent releases, i.e., little dilution of liquid wastes after discharge due to the absence of a l receiving water body. To correct the inconsistency, changes are made {

to the RSTS (3.17.1) to state that Rancho Seco will comply with the j Maximum Permissable Concentration of 10 CFR 20, Appendix B, Table II, ]

Column 2. There is no guarantee that the design objectives for 10 CFR 50, Appendix I, which is dose based, can be met singularly using concentration based LLD's, therefore Appendix I compliance statements are removed from the LCO and surveillance bases. Surveillance l requirements (Tech Spec 4.21.1) for radionuclides concentrations in i liquid effluent releases have been revised to reflect additional !

conservatism in the radioactive liquid waste sampling and analysis program by increasing the sampling and analysis frequency of dissolved and entrained gases with newly established LLD's (see Table 4.21-1). The historical mix of radionuclides released at Rancho Seco provided the basis for establishing that Cs-134 and Cs-137 are the major dose indicators for all gamma emitters, except iodines. The setting of the LLD's for Cs-134 and Cs-137 at a concentration 1 i

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 153

1. Concentration (Cont.)

equivalent of 50% of Technical Specification values has been developed and included in this amendment (see Attachment B Bases for Lower Limit of Detection Values for' Rancho Seco Liquid Effluents).

The Offsite Dose Calculation Manual (ODCM) will address dose related contributions from liquid effluents, thereby, addressing ALARA. The Offsite Dose Calculation Manual (ODCM) will be the governing document for meeting concentration standards of radioactive effluent releases.

Basis For No Significant Hazards Determination The proposed changes to the Technical Specifications regarding the concentration of radionuclides in liquid effluents do not involve a significant hazards consideration because operation of Rancho Seco in accordance with these changes would not:

(1) involve a significant increase in the probability or consequence of an accident previously evaluated. The proposed changes do not affect plant design or alter the safety / accident analysis of Chapters 11 and 14 of the Updated Safety Analysis Report (USAR). The changes provide clarification and ensure that plant liquid effluents comply with the requirements of 10 CFR 20, Appendix B, Table II, Column 2 regarding the maximum permissible concentration of radioactive materials released from the site to areas beyond the site boundary for liquid effluents. Therefore, these changes do not significantly increase the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any previously analyzed. These proposed changes reflect clarification and compliance with the requirement of 10 CFR 20, Appendix B, Table II, Column 2 and do not create the possibility of a new or different kind of accident from any previously analyzed; (3) involve a signific .nt reduction in a margin of safety. These changes ensure compliance with the requirements of 10 CFR 20, Appendix B, Table II, Column 2 and 20 CF1 20.106 regarding the maximum permissible concentration of radioactive material in the liquid effluents and the resultant exposure limit to a member of the public. Compliance with the ALARA guidelines of 10 CFR 50, Appendix I are assured by the dose limits and dose tracking methodology of the changes to Technical Specifications 3.17.2 and 4.21.2. The revision of Table 4.21-1 reflects additional conservatism in the radioactive liquid waste sampling and analysis program. The revision of definition 1.15 reflects the restructuring of the Offsite Dose Calculation Manual. Implementing procedures for dose calculation I

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FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT No. 155 PAGE 154

1. Concentration (Cont.)

will be issued in separate documents. Environmental monitoring has been removed from the ODCM, incorporated in the Radiological Environmental Monitoring Program (REMP) Manual, and added as an additional Technical Specification definition. Therefore, these revisions do not involve a significant reduction in a margin of safety.

2. Dose The existing Rancho Seco Technical Specification 3.17.2, which is provided to implement the requirements of 10 CFR Part 50, Appendix I, requires that the annual dose to a member of the public from radioactive materials in liquid effluents be limited to 3 millirems to the total body and to 10 millirems to any organ. Rancho Seco Technical Specification 3.25, which is provided to meet the dose limitations of 40 CFR 190, requires that the annual dose to a member of the public due to releases of radioactivity and radiation from fuel cycle sources be limited to 25 millirems to the total body or any organ (except the thyroid, which is limited to 75 millirems).

TheNRCpositionbstatedthatincorporationofthemodelRETSLLD values in the Rancho Seco RETS was in error since use of these values can result in releases of radioactive materials to which offsite doses may be attributed (through the use of the methodology of the Rancho Seco ODCM) that are in excess of the limits provided by the RETS to implement the regulations,10 CFR Part 50, Appendix I, and 40 CFR Part 190.

The District has developed a position with respect to the use of Lower Limits of Detection for compliance with 10 CFR Part 50, Appendix I. In this position, two sets of LLD's are used; the first is based on the capabilities of the Rancho Seco Chemistry facilities; the second is based on the capabilities of environmental-level facilities.

The historical mix of radionuclides in Rancho Seco liquid effluent was examined for dose contributions of each radionuclides. It was determined that Cs-134 and Cs-137 contributed nearly 98% of the dose due to gamma emitters to the total body and to specific organs other than the thyroid. I-131 was found to contribute essentially 100 % of the dose to thyroid. Tritium contributes variable fractions of the total dose, but is considered separately due to the distinct analysis method.

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FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENLMENT NO. 155 PAGE 155

2. Dose (Cont.)

The first set of LLD's utilizes four radionuclides as indicators for all other nuclides. Three basic cases are addressed:

1. All gamma emitters other than iodines (Cs-134 & Cs-137)

2. Iodines (I-131)

3. Tritium This set of LLD's is established at a concentration equivalent of about 50% of the values in Technical Specification 3.17.2, except for Tritium, which is at a concentration equivalent of less than 1%

(assuming an estimated annual plant effluent out flow of 20 million gallons and an average dilution flow rate of 5,000 gpm). A comprehensive system which includes administrative limits and a dose tracking program will be used to assess the cumulative offsite calculated dose with respect to the values in Technical Specification 3.17.2 prior to each release.

The second set of LLD's applies to monthly composite samples of the liquid effluent. These LLD's are established at values which represent 10% or less of the concentration equivalent of Te< tical Specification 3.17.2 (assuming an estimated annual plant ef. cent out flow of 20 million gallons and an average dilution flow rate of 5,000 gpm). The dose tracking program contains methods for updating the cumulative dose based on results of the composite sample analyses.

The mix of radionuclides in the liquid effluent will also be evaluated at semiannual intervals. (Refer to Attachment B, Bases for Lower Limit of Detection Values for Rancho Seco Liquid Effluents).

Basis For No Significant Hazards Determination The proposed changes to the Technical Specifications regarding doses due to radioactive material in liquid effluents do not involve a significant hazards consideration because operation of Rancho Seco in accordance with these changes would not:

(1) involve a significant increase in the probability or consequence of an accident previously evaluated. These changes do not significantly alter the safety / accident analysis in the Updated Safety Analysis Report (USAR). Dose and dose commitment to the Maximum Hypothetical Individual due to radioactive materials in liquid effluents are maintained to within the ALARA dose guidelines of 10 CFR 50, Appendix I. Therefore, these changes do not significantly increase the probability or consequences of an accident;

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FACILITY CHANGE SAFETY ANALYSIS LOG NO.-921 PROPOSED AMENDMENT NO. 155 PAGE 156

2. Dose (Cont.)

(2) create the possibility of a new or different kind of accident from i any previously analyzed. These changes reflect clarification in the Technical Specifications and bases for offsite dose commitment'due to plant. liquid effluents. Compliance with the ALARA dose guidelines of i 10 CFR 50, Appendix I is maintained. Therefore, these changes do not create the possibility of a new or different kind of accident; J (3) involve a significant reduction in a margin of safety. These changes provide reasonable assurance of continued compliance with the ALARA dose guidelines of 10 CFR 50, Appendix I. The dose tracking system j is required to be operated and maintained such that operating j parameters can be adjusted in accordance with the methodology in the ;

Offsite Dose Calculation Manual (ODCM). They do not allow the calculated dose values to the Maximum Hypothetical Individual, when i projected to the end of the quarter and/or year, to exceed the ALARA 1 dose guidelines of 10 CFR 50, Appendix 1. The use and definition of the Maximum Hypothetical Individual is pursuant to 10 CFR 50, i Appendix I dose methodology and is more conservative than a real dose !

to a member of the public. Therefore, these changes do not involve a significant reduction in a margin of safety.

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3. Radiological Environmental Monitoring Program I

The Rancho Seco Radiological Environmental Monitoring Program I provides for the collection and analysis of specified numbers of samples of surface water, runoff water, shoreline mud and silt, milk, fish, and several classes of harvested food at specified frequencies (Table 3.22-1). This program is based on the guidance of the model program in NUREG-0472, Rev. 2, and NUREG-0452, Rev. 5 (draft). The sampling locations are described in the REMP Manual. Table 3.22-2 provides reporting levels for radionuclides concentrations in the environmental samples in order to appropriately identify when concentrations of radioactive ma.terials and levels of radiation may be higher than expected on the basis of the effluent measurements and 1 the modeling of the environmental exposure pathways. I The changes represent conformance with the Standard RETS in NUREG-0472, Rev. 2, and NUREG-0452, Rev. 5 (draft). In addition, the radiological environmental monitoring program (REMP) will account for all potential land, water usage, and food radiological exposure pathways that exist downetream from Rancho Seco. The models (Table 3.22-1) will take account for long-term buildup of concentrations of radionuclides in bottom sediment doses due to ingesting aquatic food and direct radiation from long-term buildup of radionuclides on land 1

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a FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 157

3. . Radiological Environmental Monitoring Program (Cont.)

irrigated with contaminated water. These models are site-specific and their use is encouraged by RG 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50 Appendix I," is the basis for achieving compliance to 10 CFR 50, Appendix I. The Rancho Seco ODCM follows the guidance provided in Reg. Guide 1.109 and is now separate from the REMP.

Basis For No Significant Hazards Determination The proposed change to the radiological environmental monitoring Technical Specifications does not involve a significant hazards consideration because operation of Rancho Seco in accordance with this change would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated. These specifications provide for the measurement of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposure of individuals resulting from station operation. This supplements the radiological effluent monitoring specifications by verifying that measurable concentrations of radioactive materials and levels of radiation are not higher than expected for all potential exposure pathways. This change does not 5 alter the safety / accident analysis of Chapters 11 (11.1.7) and 14, of the USAR and therefore, does not significantly increase the probability or consequences of an accident; (2) create the possibility of a new or different kind of accident from any previously analyzed. The affected specifications concern environmental monitoring and are consistent with the guidance of Standard RETS in NUREG-0472 and NUREG-0452. Therefore, this change does not create the possibility of a new or different kind of accident; (3) involve a significant reduction in a margin of safety. The text changes represent clarification and programmatic changes to the Radiological Environmental Monitoring Program. The Radiological Environmental Monitoring Program (REMP) accounts for all significant land, water usage and food radiological exposure pathways that exist downstream from Rancho Seco. Additional sampling points, collection frequencies and reporting level requirements have been added to these specifications along with improvements to the annual land use census. Therefore, this change does not involve a significant reduction in a margin of safety.

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i FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 158

4. Land Use Census The Rancho Seco REMP, utilizing the guidance cf NUREG-0472, provides for an annual land use census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modification to the monitoring program are made if required by the results of the census. The changes here will include an addition of liquid pathway surveillance so that existing environmental and societal uses of land surrounding Rancho Seco can be kept current.

Identification of gardens in the summer, rather than the middle of winter, will be included in the census to assure a more realistic sampling of gardens. In addition, liquid and gaseous pathways are identified and reportable as land use census dose results which will be included in the Annual Radiological Environmental Operating Report.

Basis For No Significant Hazards Determination The proposed change to the Technical Specification regarding the Rancho Seco site land use census does not involve a significant hazards consideration because operation of Rancho Seco in accordance with this change would not:

(1) involve a significant increase in the probability or consequence of an accident previously evaluated. This change provides clarification and definition of site boundaries and does not significantly alter the safety / accident analysis in the Updated Safety Analysis Report.

This is consistent with Standard Tech Specs NUREG-0472; (2) create the possibility of a new or different kind of accident from any previously analyzed. This change provides an improved definition of existing geographical and site boundaries and does not create the possibility of a new or different kind of accident; (3) involve a significant reduction in a margin of safety. This change provides clarification and definition of existing plant boundaries and does not involve a significant reduction in a margin of safety.

5. Reporting, Procedures And Audits The changes to the Tech Specs 6.5.1.6, 6.5.2.8, 6.8, 6.9.2, and 6.9.5 relates to reporting, procedures and audit requirements as related to radiological effluent monitoring. These changes represent a culmination of format requirements to meet the guidelines of Standard RETS in NUREG-0472 and NUREG-0452.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 159 1

S. Reporting, Procedures And Audits (Cont.) l I

Basis For No Significant Hazards Determination f l

The proposed changes to the Technical Specifications regarding the review, ;

audit, procedures and reporting do not involve a significant hazards 4 consideration because operation of Rancho Seco in accordance with these changes would not:

I (1) involve a significant increase in the probability or consequence of j an accident previously evaluated. These changes incorporate I administrative improvements and do not alter plant design or l

,afety/ accident analysis as described in the Updated Safety Analysis Report (USAR). Therefore, these proposed changes do not significantly increase the probability of an accident; (2) create the possibility of a new or different kind of accident from previously analyzed. These proposed changes increase the scope of the administrative Technical Specifications supporting the control and the monitoring of radioactive materials in plant liquid effluents per Standard RETS in NUREG-0472 and NUREG-0452 and does not create the possibility of g a new or different accident;

'(3) involve a significant reduction in a margin of safety. These 1 administrative changes are conservative and increase the scope of the Technical Specifications for the review, audit, procedures and reporting of plant liquid effluents related activities. Therefore, these proposed changes do not constitute a significant reduction in a margin of safety.

6. Instrumentation Radioactive gaseous and liquid effluent monitoring instrumentation monitors and controls the release of radioactive materials in plant effluents. The alarm / trip setpoints for these instruments are calculated in accordance with the methodology contained in the Offsite Dose Calculation Manual (0DCM) to ensure that the limits of 10 CFR 20.106 are maintained. Changes are made to the radioactive and liquid effluent monitoring instrumentation Technical Specifications (RSTS 3.15/4.19 and 3.16/4.20, respectively) to provide clarification, editorial improvements, to establish were applicable conformance with the guidance in NUREG-0452, Rev. 5 (draf t) and NUREG-0472, Rev. 2, and to reflect current plant design and operation.

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSLD AMENDMENT No. 155 PAGE 160

6. Instrumentation (Cont.) i The existing RSTS Table 3.15-1/4.19-1 specify operability and i surveillance requirements for the Regenerant Ho'. dup c Tank (RHUT)

Discharge Line Monitor. This monitor functions as the plant liquid effluent monitor, providing automatic terminatica of plant liquid releases to ensure compliance with the limits of 10 CFR 20.106. This monitor is being replaced by a monitors downstream of the retention basins (Retention Basin Discharge Monitor) to mo re closely conform with the standard REIS. Reasonable operability and surveillance requirements are established for the RHUT total .!1ow monitor to ensure that the total volume of water released faom the A & B RHUT (effluent control point) to the Retention Basin :.s known for the determination of offsite dose. Additional changts to the bases of RSTS 3.15 reflect current plant design and operational practices regarding the processing and release of primary and secondary waste water.

1 Radioactive gaseous effluent monitoring instrumentation surveillance frequencies for the instrument channel calibratio s in Table 4.20-1 are decreased from. monthly to refueling pursuant :o the guidance contained in NUREG-0452, Rev. 5 (draft). Notations are added to ensure that the\ channel test for the Auxiliary Bu:.1 ding Stack noble gas activity monitor adequately demonstrates funct:ionality.

Allowances is provided for not performing channel testing of the d Reactor Building Purge Vent and the Auxiliary Building Stack System effluent flow rate devices when conditions pose a personnel safety hazards. These monitors are located in near proxinity to the release manifold of the Main. Steam Safety Valves and it is prudent to limit personnel access to this area during power operation.

Basis For No Significant Hazards Determination The proposed changes to the Technical Specifications regarding the radioactive gaseous and liquid effluent monitoring instrumentation do not involve a significant hazards consideration because oper.ation of Rancho Seco in accordance with these changes would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes provide clarification of alarm / trip setpoint, to ensure that the limits of 10 CFR 20.106 are maintained. The relocation of the liquid' effluent monitor allows for more comprehensive monitoring of potential radioactive effluent streams in addition to the Regenerant Holdup tanks. Therefore, these changes do not significantly increase the probability or consequence of an accident previously evaluated;

FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 161

6. Instrumentation (Cont.)

.(2) create the possibility of a new or different kind of accident from any previously analyzed. The change in location of the liquid effluent monitor allows a more comprehensive monitoring and automatic termination of potential radioactive liquid effluent streams in addition to the RHUT discharge flow path. The clarifications to the alarm / trip setpoints ensures that the limit of 10 CFR 20.106 are maintained. There are no changes to system functions and therefore-these changes do not create the possibility of a new or different kind of accident.

(3) involve a significant reduction in a margin of safety. These changes ensure that the limit of 10 CFR 20.106 are maintained. Allowances are made for instrument channel test of the Reactor Building Purge Vent and Auxiliary Building Stack System effluent flow rate devices in consideration of personnel safety during normal operation. This allowance doe not significantly reduce the reliability of the flow rate devices. Therefore, these revisions do not involve a significant reduction in a margin of safety.

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FACILITY CHANGE SAFE 1T ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 162 i Safety Analysis For Administrative And Editorial Changes (Item 48):

The changes listed in Tech Spec Item No. 48 are administrative and'

. editorial changes. They are made to improve the overall Technical Specification editorial consistency and format, clarify requirements and correct errors. Changes to Chapter 6, Administrative Controls, are made to adopt the format with Standard Tech Specs NUREG-0472 and current industry standards.

Basis For No Significant Hazards Determination:

The proposed change does not involve a significant hazards consideration because operation of Rancho Seco in accordance with this change would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated. This enhanced clarity should decrease the potential for unacceptable consequences or accidents.

These are editorial and administrative changes which do not increase the probability or consequence of an accident.

(2) create the possibility of a new or different kind of accident from any previously evaluated. A new or different kind of accident will )

-not be created d'ue to these editorial and administrative changes, i These administrative changes do not create the possibility of a new I or different kind of accident because of enhanced clarity and document consistency per Standard Tech Specs NUREG-0472.

(3) involve a significant reduction in a margin of safety. These editorial and administrative changes ensure that the Technical .

Specifications clearly address proper procedural and monitoring j control relating to radiological effluent releases and will preserve the margin of safety. Therefore, the administrative changes will not reduce the margin of safety, 1

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1 ATTACHMENT A Administrative and Editorial Changes to Rancho Seco Technical Specifications per Proposed Amendment No. 155

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Page Section 1-7 l 1.16 RESTRICTED AREA 1-7 1.17 SITE BOUNDARY 1-7 1.18 DOSE EQUIVALENT I-131 1-7 l 1.19 MEMBER (S) 0F THE PUBLIC MAXIMUM HYPOTHETICAL INDIVIDUAL 1-8 155> 1.21 1.22 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL 1-8 l 1.23 1-8 LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM 1.24 VENTILATION EXHAUST TREAlNENT SYSTEMS 1-9 1.25 PURGE'- PURGING 1-9 1.26 VENTING 1-9 1.27 RADIOACTIVE EFFLUENT 1-9 4 \

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 4

2.1 SAFETY LIMITS, REACTOR CORE 2-1 2.2 SAFETY LIMITS, REACTOR SYSTEM PRESSURE 2-4 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 2-5 3.0 LIMITING CONDITIONS FOR OPERATION 3-1 3.1 REACTOR COOLANT SYSTEM 3-1 3.1.1 Operational Components 3-1 3.1.2 Pressurization, Heatup, and Cooldown Limitations 3-3 3.L 3 Minimum Conditions for Critica11ty 3-6 3.1.4 Reactor Coolant System Activity 3-8 3.1.5 Chen:istry 3-10 Proposed Amendment No.155 ifa a

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h RANCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS i TABLE OF CONTENTS (Continued) i i

S Page

-.e. c ti on 3.14.5 Fire Hose Stations 3-57 ,

3.14.6 Fire Barrier Penetration Fire Seals 3-58 ;

155>< 3.15 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3-60 3.16 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 3-63 3.17 LIQUID EFFLUENTS 3-70 3.17.1 Concentration 3-70 3.17.2 Dose 3-71 3.17.3 Liquid Holdup Tanks 3-72 155>< 3.17.4 Liquid Effluent Radwaste Treatment 3-72a 3.18 GASE0US EFFLUENTS 3-73 3.18.1 Dose Rate 3-73 l 155> 3.18.2 Dose-Noble Gases 3-74 3.18.3 Dose-Iodine-131, Iodine-133, Tritium, and Radioactive Materials in Particulate Form 3-75 3.18.4 Gaseous Radwaste Treatment 3-78 3.18.5 Gas Storage Tanks 3-79 3.19 DELETED l t

< 3.20' DELETED i 3.21 SOLID RADI0 ACTIVE WASTES 3-80 ;

3.22 RADIOLOGICAL ENVIRONMENTAL MONITORING 3-81 3.23 LAND USE CENSUS 3-87 l 3.24 EXPLOSIVE GAS MIXTURE 3-89 3.25 FUEL CYCLE DOSE 3-90 3.26 INTERLABORATORY COMPARISON PROGRN4 3-92 Proposed Amendment No.155 i

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RANCHO SECO UNIT 1

. TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Page !

Section 4.14 SHOCK SUPPRESSORS (SNUBBERS) 4-47 4.15 RADI0 ACTIVE MATERIALS SOURCES 4-48 Reserved 4-49 4.16 4.17 STIAMGENERATORS 4-51 4.17.1 Steam Generator Sample Selection and Inspection 4-51 4.17.2 Steam Generator Tube Sample Selection and Inspection 4-51 4.17.3 Inspection Frequencies 4-52 4.17.4 Acceptance Criteria 4-53 4.17.5 Reports 4-54 4.17.6 OTSG Auxiliary Feedwater Header Surveillance 4-54 4.17.7 Inspection Acceptance Criteria and Corrective Actions 4-55 4.17.8 Reports 4-55 4.18 FIRE SUPPRESSION SYSTEM SURVEILLANCE 4-58

'155>< 4.19 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 4-63 4.20 RADIOACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 4-65 4.21 LIQUID EFFLUENTS 4-69 4.21.1 Concentration 4-69 155>< 4.21.2 Doses 4-72 4.21.3 Liquid Holdup Tanks 4-73 155> 4.21.4 Liquid Effluent Radwaste Treatment 4-73a Proposed Amendment No. 155 vi 1

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TECHNICAL SPECIFICATIONS j 1

TABLE OF CONTENTS (Continued) {

I Section Page 155> 4.22 GASEOUS EFFLUENTS 4-74 4.22.1 Dose Rate 4-74 4.22.2 Dose-Noble Gases 4-77 4.22.3 Dose-Iodine-131, Iodine-133, Tritium, and Radioactive Materials in Particulate Form 4-79 Gaseous Radwaste Treatment 4-79 4.22.4 i Gas Storage Tanks 4-80 4.22.5 Deleted 4-79 !

4.23 4.24 Deleted 4-80 4-81

< 4.25 SOLID RADI0 ACTIVE WASTES 4.26 RADIOLOGICAL ENVIRONMENTAL MONITORING 4-83

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4-86 4.27 LAND USE CENSUS I

4.28 EXPLOSIVE GAS MIXTURE 4-87 4 4.29 FUEL CYCLE DOSE 4-89 4.30 INTERLABORATORY COMPARIS0N PROGRN4 SURVEILLANCE REQUIREMENT 4-90 4.31 NUCLEAR SERVICE ELECTRICAL BUILDING EMERGENCY HEATING 4-91 VENTILATION AND AIR CONDITIONING 5.0 DESIGN FEATURES 5-1 I

5.1 SITE 5-1 155> 5.1.1 Exclusion Area 5-1 5.1.2 Low Population Zone 5-1 5.1.3 Site Boundary for Gaseous Effluents 5-1 5.1.4 Site Boundary for Liquid Effluents 5-1 4

Proposed Amendment No. 155 vii

155>

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS

' TABLE OF CONTENTS (Continued) ,

Page Section {

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5-2 5.2 CONTAINMENT 5.2.1 Reactor Building- 5-2 ]

5.2.2, Reactor Building Isolation System 5-3 5.3 REACTOR 5-4 Reactor Core 5-4 5.3.1 5.3.2 Reactor Coolant System 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 New Fuel Inspection and Temporary Storage Rack 5-6 5.4.2 Neb and Spent Fuel Storage Racks and Failed 5-6 Fuel Storage Container Ract.[

5.4.3 New and Spent Fuel Temporary Storage 5-6 5.4.4 . Spent Fbel Pool and Storage Rack Design -

5-6 ti 1

Proposed Amendment No. 155 -

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS i TABLE OF CONTENTS (Continued)

Page Section 6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 ORGANIZATION 6-1 6.2 6.3 FACILITY STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND AUDIT 6-3 6.5.1 Plant Review Committee 6-3 6.5.2 Management Safety Review Committee 6-6 6.5.3 Technical Review and Control 6-8 138>

4 6.5.4 Audits 6-9 6 . 16 REPORTABLE OCCURRENCE ACTION 6-10 6.7 SAFETY LIMIT VIOLATION 6-11 n

6.8 PROCEDURES 6-11 6.9 REPORTING REQUIREMENTS 6-12 6.10 RECORD RETENTION 6-13 6.11 . RADIATION PROTECTION PROGRAM 6-14 6.12 RESPIRATORY PROTECTION PROGRAM - Deleted 6.13 HIGH RADIATION AREA 6-15 6.14 ENVIRONMENTAL QUALIFICATION 6-16 6.15 PROCESS CONTROL PROGRAM (PCP) 6-17 155> 6.16 0FFSITE DOSE CALCULATION AND RADIOLOGICAL MONITORING 6-18

< PRDGRAM MANUAL 5 6.17 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (LIQUID, GASEOUS, AND SOLID) 6-19 6.18 POSTACCIDENT SAMPLING 6-22 Proposed Amendment No.'138, Rev. 2 Proposed Amendment No.:155 vii3

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES Page Section '

1-10 155H 1.9-1 FREQUENCY NOTATION 2-9 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 3-27 3.S.1-1 INSTRUMENTS OPERATING CONDITIONS 3.5.5-1 ACCIDENT MONITORING INSTRUMENTATION OPERABILITY REQUIREMENTS 3-38b 3-40 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP YALUES 3-4.la 3.7-2 VOLTAGs PROTECTION SYSTEM LIMITING CONDITIONS 3-41b 3.12-1 SAFETY RELATED HYDRAULIC SNUB 8ERS 3-51a-e 3.14-1 FIRE DETECTION INSTRUMENTS FOR' SAFETY SYSTEMS 3-55' 3.14-1 INSIDE BUILDING FIRE HOSE STATIONS 3-57a 3.15-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3-61 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 3-64 A 3-83

. 3.22-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 2.22-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS 3-86 IN ENVIRONMENTAL SAMPLES 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS 4-3 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY 4-8 4.1-3 HINIMUM SAMPLING FREQUENCY 4-9 4.2-1 CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 4-12b l 1

4.10-1 ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-42 4.10-2 OPERATIONAL ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-22a l

Proposed Amendment No. 155 iX

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-TECHNIC.".L SPECIFICATIONS LIST OF TABLES (Continued) l

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Page Section 4-47d,e 4.14-1 DESIGNATED SAFETY RELATED HYDRAULIC SNUBBERS FUNCTIONALLY TESTED ONLY AS REQUIRED BY THE SNUBBER SEAL REPLACEMENT PROGRAM 4-56 4.17-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED 00 RING INSERVICE INSPECTION 4-57 4.17-2A STEAM GENERATOR TUBE INSPECTION STEAM GENERATOR TUBE INSPECTION (SPECIAL LIMITED AREA) 4-57a 4.17-2B 4.17-3 OTSG AUXILIARY FEEDWATER HEATER SURVEILLANCE 4-57b,c 4-64 4.19-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.20-1 RAb!0ACTIVEGASE0USEFFLUENTMONITORINGINSTRUMENTATION 4-66 SURVEILLANCE REQUIREMENTS 4.21-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM 4-70 4.22-1 RADI0ACkIVEGASEOUSWASTESAMPLINGANDANALYSISPROGRAM 4-75 4-84 it- 4.26-1 MAXIMUM VALUES.FOR THE LOWER LIMITS OF DETECTION (LLD) 4.28-1 EXPLOSIVE GAS MIXTURE INSTRUMENTATION SURVEILLANCE 4-88 REQUIREMENTS ,

MINIMUM SHIFT CREW COMPOSITION 6-2 138>< 6.2-1 .

Pr> posed Amendment No. 138, Rev. 2 Pioposed Amendment No. 155 X j

RANCHO SECO UNIT 1 ')

1 TECHNICAL SPECIFICATIONS l

I LIST OF FIGURES (Continued)

Figure 3.5.2-7 Core Imbalance vs. Power Level, O to 40 EFPD 3.5.2-8 Core Imbalance vs. Power Level, after 30 EFPD 4

3.5.2-9 Core Imbalance vs. , Power Level, after 300 EFPD with APSRs Withdrawn 3.5.2-10 Deleted 3.5.2-11 ' Deleted 3.5.2-12 Deleted 3.5.4-1 ' Incore Instrumentation Specification Axial Imbalance Indication ,

3. 5.4-2 Incore Instrumentation Specification Radial Flux Tilt Indication

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3.5.4-3 Incore Instrumentation Specification 3.18-1 General Layout of Site 4.13-1 Main Steam Inservice Inspection <

4.13-2 Main Feedwater Inservice Inspection

! 4.13-3 Main Steam Dump Inservice Inspection 155> 5.1-1 Exclusion Area 5.1-2 Low Population Zone 5.1-3 Site Boundary for Gaseous Effluents

< 5.1-4 Site Boundary for Liquid Effluents 138> 6.2-1 SMUD Corporate Support of Nuclear Safety to Rancho Seco

< 6.2-2 Nuclear Organization Chart l

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Proposed Amendment No. 138, Rev. 2 Proposed Amendment No. 155 xii

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TABLE l'.9-1 ]

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

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

i W At least once per 7 days. l l M At least once per 31 days.

Q At least once per 92 days.

SY At least once per 184 days.

A At least once per 12 months R At least once per 18 months.

BA At least once per 24 months S/U Prior to each reactor startup.

P Completed prior to each release.

NA Nop appifcable.

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i Proposed Amendment No. 155 1-10 4

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls RESPONSIBILITIES (Continued)

h. Perfomance of special reviews and investigations and reports l

138>< thereon as requested by the AGM, Nuclear Power Production.

138> 1. Review of the Plant Security Plan and changes thereto.

j. Review of the Emehjency Plan and changes thereto.

k. Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE 155> DOSE CALCULATION MANUAL and the RADIOLOGICAL ENVIRONMENTAL l

MONITORING PROGRAM MANUAL. (See Specifications 6.15 and 6.16.)

1 Review of major changes to the Radioactive Waste Treatment Systems (Liquid, Gaseous and Solid), and all information required l < by Specification 6.17.

155> m. Review of any accidental, unplanned, or uncontrolled release of radioactive material to the environs including the preparation and 3 I

forwarding of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence, anu the forwarding of these reports to the Nuclear Plant Manager and to the

< MSRC.

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AUTHORITY i

6.5.1.7 The Plant Review Comittee shall:

138>< a. Recomend in writing to the AGM, Nuclear Power Production approval or disapproval of items considered under 6.5.1.6(a) through !

155>< (m) above. !

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b. Render determinations in writing with regard to whether or not i 155>< each item considered under 6.5.1.6(a) through (e), and (1) above ]

constitutes an unreviewed safety question.

c. Provide imediate written notification to the Chaiman of the Management Safety Review Comittee of disagreement between the 138> PRC and the AGM, Nuclear Power Production; however, the AGM, 4 Nuclear Power Production shall have responsibility for resolution of such disagreements pursuant to 6.5.1.1 above.

RECORDS 6.5.1.8 The Plant Review Comittee shall maintain written minutes of each 138> meeting and copies shall be provided to the AGM, Nuclear Power

< Production and the Chaiman of the Management Safety Review Comittee.

Proposed Amendment No. 138, Rev. 2 Proposed Amendment No. 155 6-5

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 138>< 6.5.4 (Continued)

a. The conformance of facility o'peration to all provisions contained within the Technical Specifications and applicable license conditions at least once per year.

b. The performance, training and qualifications of the District's entire facility technical staff at least once per year.

c. The result of all actions taken to correct deficiencies occurring in facility equipment, structures, systems or methods of operation that affect nuclear safety at least once per six 138,4 months for those changes not previously audited.

d. The performance of all acti'tities required by the Quality Assurance Program to meet the criterf a of Appendix "B",10 CFR 50, at least once per two (2) years.

e. The Facility Emergency Plan and implementing procedures at least ";

once per two (2) years.

t. The Facility Security Plan and implementing procedures at least once per two (2) years.

g. Any other area of facility operation considered appropriate by 138>< the MSRC, Deputy General Manager, Nuclear or the General Manager. 1

h. Complia6ce with fire protection requirements and implementing procedures at least once per two (2) years.

d

i. An independent fire protection and loss prevention inspection and 1 audit shall be performed annually utilizing either qualified offsite licensee persennel or an outside fire protection firm.

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j. An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than three (3) years.

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

1. The 0FFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

155> m. The PROCESS CONTROL PROGRAM and implementing procedures for

< processing and packaging of radioactive wastes from liquid systens at least once per 24 months.

155> n. The performance of activities required by the Quality Assurance

< Program for Effluent Control and Environmental Monitoring.

138> Audit reports of reviews encompassed oy Section 6.5.4 shall be forwarded to the General Manager,11SRC Chairuan, and to the management positions responsible for the areas reviewed within

< 30 davs af ter ccnoletion.

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ,j 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 provisions of 10 CFR 50.36 (c) (1) (1) and 10 CFR 50.72 shall {

be complied with.

138> b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Director, Nuclear Operations and Maintenance, the AGM, Nuclear Power Production, and the Chairman of the MSRC shall be notified 4 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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

d. The Safety Limit Violation Report shall be submitted to the 138> Commission, the MSRC, and the AGM, Nuclear Power Production,

< within 14 days of the violation.

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

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation. j

f. Fire Protection Procedures implementation. i 155> g. PROCESS CONTROL PROGRAM implementation, l

h. OFFSITE DOSE CALCULATION MANUAL implementation. j

i. RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL l l

implementation.

j. Quality Assurance Program for the' Effluent Contrel and Environmental Monitoring using the guidance of Regulatory Guide

< 4.15, Revision 1. February 1979.

138> 6.8.2 Each procedure of 6.8.1 above and changes thereto shall be reviewed

< and approved as set forth in Specification 6.5.

Proposed Amendment No.138, Rev. 2 Proposed Amendment No. 155 6-11 L

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.8 PROCEDURES (Continued) 6.8.3 Temporary changes to procedure 6.8.1 above may be made provided:

-a. The intent of tne original procedure is not altered.

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

c. The change. is documented, reviewed by the PRC and approved by the Plant Superintendent within seven (7) days of implemen-tation.

6.9 REPORTING REQUIREMENTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

Startup Report 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitt following (1) Receipt of an operating license; (2) amendment o'f the license involving a planned increase in power level; (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier; and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and comparison of these values with design predictions and specifications. Any 1 corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

6.9.1.2 Startup reports shall be submitted within (1) Ninety (90) days following completion of the startup test program; (2) Ninety (90) days following resumption or commencement of commercial power operation; or (3) Nine (9) months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test r program and resumption or commencement of commercial power l operation), supplementary reports shall be submitted at least every three (3) months until all three events have been completed.

Proposed Amendment No. 155 6-12

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 155>< 6.9.2 Radiological Reoorts 6.9.2.1 Annual Radiological Reports Annual reports covering the activities of the unit, as described 155> below, for the previous calendar year shall be submitted as follows:

6.9.2.1.1 Annual Occupational Radiation Exposure Report The Annual Occupational Radiation Exposure Report shall be submitted to the Commission within the first calendar quarter of each calendar year in compliance with 10CFR20.407.

6.9.2.1.2 Annual Exposure Report The Annual Exposure Report shall be submitted to the Commission within the first calendar quarte'r of each calendar year in

< accordance with the guidance contained in Regulatory Guide 1.16.

6.9.2.2 Annual Radiological Environmental Operating Report 155>< 6.9.2.2.1 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar 155>< year hhall be ' submitted prior to May 1 of each year.

g' 155>< 6.9.2.2.2 The Annual Radiological Environmental Operating Reports shall include sunmaries, 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 surveillance reports, and an assessment of the observed impacts of the plant operation on the envi ronment. The reports shall also include the results of the 155<> Land Use Census required by Specification 3.23. 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.

155><

Proposed Amendment No. 155 6-12a i, .

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RANCHO'SECOUNIT1 TECHNICAL SPECIFICATIONS

. Administrative Controls I

.6.9.2.2.2 .(Continued) .

J 155>< The Annual Radiological Environmental Operating Repor'ts shall l 138> include summarized and tabulated results of all radiological environmental , samples taken during the report period. In the

, event .that some re' s ults are not 'available.for inclusion.with the report, the report shall be, submitted noting and explaining the-reasons for 'the missing r'esults. The. missing data shall be' sub~mitted as soon as possible in a supplementary report.

The reports shall also include the following: a. summary

, 155> description of the Radiological Environmental Monitoring .

< ' Program; including sampling methods for each sainple type,. size and physical characteristics of.each sample type, sample preparation methods, analytical ' methods, and measuring equipment usedL a map of all sampling locations keyed to a table giving -

dist'ances and directions from one reactor; the result'of land ~

use censuses,. and the results of licensee participation in the Interlab" Comparison Program. .The annual . report shall also .

155> iriclude.informa. tion related to Specificati'on 4.29, Uranium Fuel

< Cycle Dose. .

6.9.'2.3. Semiannual' Radioa' c tive ' Effluent Release Report 155><' . Routine Se'miannual Radioactive Effluent Release Repor'ts covering.the operatioit of/the unit .during 'the previou.s .six moriths pf operation shall be. submitted'within 60 days af ter January 1 and July' 1 of each -

year. - - -

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155>< 6.9.2.3.1. The Semiannual Radioactive Effluent Release Reports shall include a summary of the ' quantities of radioactive liquid and gaseous effluents and solid waste released f rom-the unit as outlined in Regulatory Guide 1.21, "He'asuring, Evaluating, and Reporting Radioactivity in Solid Wastes /and Releases of

. Radioactive. Materials in Liquid' and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized i on a quarterly basis, following the format of Appendix B thereof.

155> The Semiannual Radioa.ctive Effluent Release Report shall include a summary of hourly meteorological data collected over the

< report period.

Proposed Amendment No. 138, Rev. 2 Proposed Amendment No.155 6-12b

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2.3.1 Continued) 1554> The Semiannual Radioactive Effluent Release Reports shall include an assessment of the radiation doses from radioactive gaseous and liquid effluents to individuals due to their 1554> activities inside the site boundary (Figures 5.1-3 and 5.1-4) during the report period. All assumptions used in making these assessments (e.g., specific activity, exposure time, and location) shall be included in these reports.

1554> The Semiannual Radioactive Effluent Release Reports shall include the following infomation for all unplanned releases to unrestricted areas of radioactive materials in gaseous and liquid effluents:

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 radioactive effluent release reports shall include an assesshent of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar .

1 quarter, as outlined in Regulatory Guide 1.21. The assessment of radiation doses shall be perfomed in accordance with the 155* OFFSITE 00SE' CALCULATION MANUAL (ODCM).

The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP), RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL and 0FFSITE DOSE CALCULATION MANUAL (0DCM) pursuant to Specifications 6.15 and 6.16 as well as any ,

major changes to Liquid, Gaseous or Solid Radwater Treatment i Systems pursuant to Specification 6.17.

I The Semiannual Radioactive Effluent Release Report shall include tables for comparison with Specifications 3.17.2, 3.18.2, and 3.18.3. The July-December report shall include a summary table -

for comparison with the annual values in Specifications 3.17.2, 3.18.2, and 3.18.3.

The Semiannual Radioactive Effluent Release Report shall also 3 include events described in Specifications 3.17.1, 3.17.3,

( 3.18.1 and 3.20.

Proposed Amendment No. 155 6-12c 4

. RANCHO SECO UNIT'l-TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2.3.1 Continued)

'155> The Semiannual Radioactive Effluent Release Report shall include the following information for each type of solid waste shipped offsite during the report period:

a. Container volume,

b. Total curie quantity (determined by measurement or estimate),

c. Principal radionuclides (determined by measurement or estimate),

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

e. Type of ' container (e.g., LSA, Type A, Type B, High Integrity), and

f. Solidification agent (e.g., cement).

MONTHLY REPORT 6.9.3 Routine rehorts of operating statistics, including narrative summary ,

of operating and shutdown experience., of lifts of the Primary System I q- Safety Valves or.EMOVs, of major safety related maintenance, and tabulations of facility changes, tests or experiments required pursuant to 10 CFR 50.59(b), shall be submitted on a monthly basis to the Office of Director of Inspection and Enforcement, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, with a copy to the !

Regional Office, postmarked no later than the 15th day of each month following the calendar month covered by the report.

LICENSEE EVENT REPORT 6.9.4 The LICENSEE EVENT REPORTS of Specification 6.9.4.1 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC as Licensee Event Reports. Supplemental reports -

may be required to fully describe final resolution.of occurrence. In case of corrected or suppler, ental reports, a , License Event Report shall be completed and reference shall be made to the original report date, pursuant to the requirements of 10 CFR 50.73.

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Proposed Amendment No. 155 6-12d

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j l4 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS !

Administrative Controls l

LICENSEE EVENT REPORT 6.9.4.1 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty (30) drJs of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report fom.

~ l pursuant to 10 CFR 50.73 and the guidance of NUREG-1022. :

a. (i) The completion of any nuclear plant shutdown required by the plant's Technical Specification; or (ii) any operation or condition prohibited by the plant's

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Technical Specifications; or (iii) Any deviation from the plant's Technical Specifications authorized pursuant to 10 CFR 50.54(x).

b. Any event or condition that resulted in the condition of tiie nuclear power plant, including its principal safety barriers, being seriously degraded, or that resulted in the nuclear power plant being:

(1) In an unanalyzed condition that significantly compromised plant safety; (ii) iN a condition that was outside the design basis of the plant; or

In a condition not covered by the plant's operating (iii) and emergency procedures.

c. any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.

d. Any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (RPS). However, actuation of an ESF, including the RPS, that resulted from and was part of the preplanned sequence during testing or reactor operation need not be reported. :

e. any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to:

1. Shut down the reactor and maintain it in a safe shutdown condition; Proposed Amendment No.155 155>< e 6-12e

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls LICENSEE EVENT REPORT

2. Remove residual heat;

3. Control the release of radioactive material; or

4. Mitigate the consequences of an accident. ,

f. Events covered in paragraph 6.9.4.1.e of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to this paragraph if redundant equipment in the same system was operable and available to perfonn the required safety function.

g. Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

1. Shut down the reactor and maintain it in a safe shutdown condition;

2. Remove residual hea,t;-

g

3. Control. the release of radioactive material; or

4. Mitigate the consequences of an acciden't.

h. 1. Any airborne radioactivity release that exceeded 2 times the applicable concentrations of the limits specified in Appendix B. Table II of 10 CFR 20 in unrestricted areas, when averaged over a time period of one hour.

2. Any liquid effluent release that exceeded 2 times the limiting combined Maximum Permissible Concentration (MPC)

(see Note 1 of Appendix B to 10 CFR 20) at the point of entry into the receiving water (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases, when averaged over a time period of one hour.

i. Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.

j. Failure of the pressurizer EMOVs or Primary System Safety Valves.

Proposed Amendment No.155 e

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls Special Reports 6.9.5 Special reports shall be submitted to the Regional Administrator, Region V 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:

i A. A one-time only, " Narrative Sumary of Operating Experience" l will be submitted to cover the transition period (calendar year j 1977). l B. A Reactor Building Structural integrity report shall be submitted within ninety (90) days of completion of each of the following tests covered by Technical Specification 4.4.2 (the integrated leak rate test is covered in Technical Specification 4.4.1.1). -

1. Annual Inspection

2. Tendon Stress Surveillance

3. End Anchorage Concrete Surveillance

4. Liner Plate Surveillance C. Inserhice Inspection Program 4 D. Inoperable Accident Monitoring Instrumentation 30 days (3.5.5)

E. Status of Inoperable Fire Protection Equipment F. Inoperable Emergency Control Room /TSC Ventilation Room Filter System G. Radioactive Liquid Effluent Dose 30 days (3.17.2)

H. Noble Gas Limits 30 days (3.18.2)

I. Radiofodine and Particulate 30 days (3.18.3) 155> J. Gaseous and Liquid Radwaste Treatment 30 days (3.18.4 and 3.17.4)

< K. Radiological Environmental Monitoring Program 30 days (3.22)

L. Monitoring Point Substitutions 30 days (3.22) 155>< M. Solid Radioactive Wastes 30 days (3.21) '

N. Fuel Cycle Dose 30 days (3.25) 155>< 0. Land Use Census 30 days (3.23)

P. Steam Generator Tube Inspection 30 days (4.17.5)

Proposed Amendment No. 155 155>< 6-12g r __ _- ..

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I I RANCHO SECO UNIT 1-TECHNICAL SPECIFICATIONS Administrative Controls l 6.15 PROCESS CONTROL PROGRAM (PCP) 6.15.1 Function 155> The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and 1 I

formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured.

i 6.15.2 Changes A. The PCP shall be approved by the Commission prior to implementation.

B. Licensee initiated changes to the PCP shall:

1. Be submitted to the Commission by inclusion in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was/were made and shall contain:

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

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b. A determination that the change did not reduce the 3

overall conformance of the solidified waste product to 1 existing criteria for solid wastes; and

c. Documentation of the fact that the change has been reviewed and found acceptable by the Plant Review Committee.

2. Become effective upon review and acceptance by the PRC, unless otherwise acted upon by the Commission through written notification to the Licensee.

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Proposed Amendment No.155 6-17 1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 155>< 6.16 0FFSITE DOSE CALCULATION AND RADIOLOGICAL ENVIRONMENTAL MONITORING FROGRAM MANUAlb 6.16.1 Function 155> 6.16.1.1 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall describe the methodology and parameters to be used in the calculation of offsite

< doses due to the release of radioactive material in gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable LCO's contained in these Technical Specifications. Methodologies and calculational procedures acceptable to the Commission are contained in various Regulatory Guides as noted in the bases of applicable LCO's.

6.16.1.2 The RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) MANUAL shall be a manual containing the description of the Rancho Seco Radiological Environmental Monitoring Program. The REMP manual shall contain a description of the environmental samples to be collected, the sample locations, sampling frequencies, and sample analysis criteria.

6.16.2 Any changes to the ODCM or REMP MANUAL shall be made as follows:

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A. Licensee-initiated changes:

1. Shall be submitted to the Commission by inclusion in the 155> Semiannual' Radioactive Effluent Release Report for the 4 period in which the change (s) was/were made and shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist' of a package of those pages 155><

of the ODCM and the REMP MANUAL 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 PRC and MSRC.

2. Shall becore effective upon a date specified and agreed to by both the PRC and MSRC following their review and acceptance of the change.

Proposed Amendment No.155 6-18 e _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___

RANCHO SECO UNIT 1 TECHNICAL SPEC 1FICAT10NS Administrative Controls l

6.17 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (LIQUID, GASEOUS, ANU SOLID) 6.17.1 Function The radioactive waste treatment system (liquid, gaseous, and solid) 155>< are those systems described in the facility Updated 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 LCOs set forth in these Specifications.

Major changes to the radioactive waste systems (liquid, gaseous, and I 6.17.2

^

138>

sol-i'd) .shall be made by the following method. For 'the~ purpose of t.his I

< specification, " major changes' is defined in Specification 6.17.3.

Licensee-initiated changes: f

.138> j

1. The Commission shall be informed of all changes by the inclusion j of a suitable discussion of each change in the Semiannual l Radioactive Release Report for the period in which the changes l l

were made. The discussion of each change shall contain:

a. A summary of the evaluation that led to the determination thct the change could be made in accordance with 10 CFR 50.59;

b. Sufficient infomation to support the reason for the change 9

without benefit of additional or supplemental information;

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

d. An evaluation of the change with regard to the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste if different from those previously predicted in the license application and amendments thereto; I 1

155> e. An evaluation of the change with regard to the expected f maximum exposures to a MEMBER OF THE PUBLIC in the j UNRESTRICTED AREA and to the general population if different q from those previously estimated in the license application i and amendments thereto; J i

Proposed Amendment No. 138, Rev. 2 Proposed Amendment No.155 6-19 k ___ _ _ _ _ _ _ _ _ _ _

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. RANCHO SECO UNIT 1 TECHNICAL' SPECIFICATIONS TABLE OF CONTENTS APPENDIX B PAGE 1.0 DEFINITIONS l' '

2.0' 4 ENVIRONMENTAL PROTECTION CONDITIONS l

.2.1 Deleted 2.2' Deleted 2.3 Deleted 2.4 Deleted- '

2.5 Deleted 155><. 2.6 Deleted 20 i

25 3.0 NON-RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE PRDGRAMS 3.1 Erosion 25 3.2 Drift Contaminants 25 3.3 Deleted 3.4 Noise 26 3.5 Fogging 27 3.6 Deleted

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4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 31 4.1 Deleted 4.2 Deleted 4.3 Deleted'

4. 4. Deleted 4.5 Deleted 4.6 Deleted 4.7 Deleted 4.8 Deleted 4.9 . Deleted 4.10 Deleted i 5.0 ADMINISTRATIVE CONTROLS 42 5.1 Responsibility 42

' 5.2 Organization ' 42 5.3 Review and Audit 42 5.4 Action to be taken in Event of Violation of an Environmental Protection Limit 43 i 5.5 Procedures 44 5.6 Plant Reporting Requirements 44 5.7 Records Retention 46 5.8 Deleted Proposed Amendment 155 ,

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\ THIS PAGE HAS BEEN DELETED i

4 Proposed Amendment No.155 .

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Proposed Amendment No.155 j J

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ATTACHMENT B Bases for Lower Limit of Detection Values for Rancho Seco Liquid Effluents

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CASED FOR LOWER LIMIT OF DETECTION VALUE3 l FOR RANCHO SECO LIQUID EFFLUENTS l

I 1&q INTRODUCTION In a previous transmittal .16 to USNRC Region V, a statist-ical basis for lower limit of detection (LLD) of radio-nuclide concentrations.in Rancho Seco effluents was estab- .]

lished. As indicated in that document, as the background 1

for the analysis,of interest is lowered, the ratio of the {

dose represented by the LLD to the Technical Specification value also lowers. (Fig. 1) In the limit, gg the background approaches 0, the ratio approaches 0.5.

The Sacramento Municipal Utility District proposes in this document a program which utilizes LLD's for the the liquid effluents at values which result in this ratio of 921 This I

program features a comprehensive management control and oversight process and dose tracking system. This approach is considered a,ppropriate due to the unusual characteristics of the Rancho Seco site in which the liquid outfall of the site is the headwater of Clay Creek. Also, during much of the year, this flow also is a large part of the flow of Hadselville and Laguna Creeks.

The objective of this program is to achieve and maintain compliance to 10 CFR Part 50, Appendix I in a manner specific to the unusual environmental setting of the Rancho Seco site.

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,2Ag HISTORY The original Technical Specifications for liquid effluents were based on two' premises: (Fig. 2)

(1) The concentrations'in 10 CFR Part 20 applied;

-(2) Normal operation precludedfradioact'ivityLin the secondary' system, and thusLin the~ liquid' effluents.

However, steam generator. tube leaks developed. Radioact-ivity began to appear frequently in the liquid effluent in 1981.

In 1984, liquid effluent Technical Specifications were adopted which contained LLD values-which corresponded' directly-to the values used for the " standard plant". These i values were used \ until it was determined that the calculated-1 doses associated with these values were somewhat in excess of the Technical Specification values. (Fig. 3) 'At that time, interim LLD's were adopted to bring the situation more in agreement with Appendix I. (Fig. 4)

Usage factors on the stream system also changed in 1985.

Prior to-1985, the stream system was extensively used by area' residents downstream from the site boundary fence.-

Individuals claimed to have large consumption of aquatic foods from this area. After 1984, information from these individuals indicated that usage of the stream system upstream from Laguna creek has essentially. stopped. In addition, the area shown in Figure 5 was posted as a "No 2

Trespassing, No Hunting, No Fishing" area.

During the period 1981-84, a database was established for the quantities of radionuclides released in the liquid effluent. The data for gamma emitters are presented in Table 1. Note that cs134 and cs137 together comprise nearly half of the activity. These values are used as the histor-ical radionuclides mix for dose and LLD calculations.

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222 DOSE AED CONCENTRATION CALCULATIONS 3.1 Dose Calculation for Historical Mix )

The projected dose calculation methodology used at Rancho j Seco is based on Regulatory Guide 1.109 as implemented by 1

the LADTAP software. This methodology was used to determine the fraction of total dose associated with each. radionuclides in the historical mix for the total body and for the most limiting organ for the nuclide. It was found that virtually all dose to the thyroid is due to iodines, so the contri-bution from iodines was separated from the rest. The results are presented in Table 2. This table clearly indicates that Cs134 and Cs137 are by far the major contri- l l

butors to dose at 97.8% of the total body dose and 97.6% of the limiting organ dose. (Except thyroid)

\ 1 The FIRST PROPOSAL is that Cs134 and Cs137 will be indicators for dose for all gamma-emitting radionuclides except iodines. Table 3 contains the resulting dose fractions.

The usage factors established before 1985 will continue to be used to provide conservatism until the land use census currently underway has been completed and the data evaluated.

3.2 Concentration (LLD) Calculations .

In the methodology as implemented by LADTAP, dose is propor-tional to the total quantity of activity released, all other factors being constant. Three of these factors are under operational control. (Fig. 6) The others are built in to the database for the Offsite Dose Calculation Manual. (ODCM) 4

,. ' r The;three factors under operational control include: (Fig.

6)

(1) Annual volume of liquid effluent released from Regenerant Hold-Up Tanks A & B (RKUT A & B)

(2) Concentration of radioactive materials in the effluent (3)~ Flow rate at'the site outfall.

The flow rate at the site outfall is normally held constant at 5000 i 10% gpm.. The concentrations in question are the LLD's to be establishe'.d The annual volume through RHUT's A-

& B is also considered to be~ constant at 20 million gallons.

With cs134 and cs137 as the dose indicators for all gamma emitters, except iodines, the concentration equivalent of dose at the Technical Specification values of 3 mrem total 1

. body and 10 mres organ is illustrated in Figure 7.

The SECOND PROPOSAL is to place the LLD equivalent of dose ,

at 50% of the Technical Specification values. (Fig. 7) 3.3 Laboratory Significance The proposed LLD's are achievable with a modest upgrade in equipment in the facilities which are available. Counting t'mes i of 4000-6000 seconds are expected. State-of-the-art j equipment and specialized facilities; are 'ng_t; required to support the plant with reasonable delays.

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.M MANAGEMENT AND ADMINISTRATIVE CONTROLS 44.1- Management Control System ;

The effluent control 1 program has been in a continuous state of evaluation and improvement since late 1985. One result ,

has been to develop a comprehensive Offsite Exposure Control Program. (OECP) This' program defines policy, responsi- l

.I bilities, and interfaces. ')

The OECP is deriv.ed from a foundation of Technical Specifi- '

cations, Regulatory Guides, the Safety Analysis Report, Industry Standards, and Standard District Instructions.

(Fig. 8) It addresses all effluent paths (gaseous, liquid, 't solid) , below regulatory concern, and related programs such as meteorology.

The policy document (Fig. 8) is the Offsite Exposure Control i Management Process.- It defines the functional areas which' are subject to the OECP and the responsibilities'of'each 1 organization which is linked to the Program. Indirect i linkages, such as Training Department, are included.

Responsibility for the Program at Rancho Seco is placed on j the Deputy General Manager, Nuclear Organization. l l

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The implementing document (Fig. 8) . for the Management Process is the Offsite Exposure Control Manual. This manual addresses the following:

(1) OECP goals and objectives (2) Measures of organizational performance, management involvement, and Station trends 6

Functional and organizational interfaces j

. (3).

(4) Definition of data and information management systems. ;

I As indicated, (Fig. 8) the Offsite Exposure Control Manual provides the interfaces among effluent control, offsite dose calculation, environmental monitoring, solid radwaste i process control, engineering, and other related programs.

The ALARA Manual defines ALARA goals and objectives for all .

i Rancho Seco functions, including the OECP.

4.2 Administrative and Operational Controls These controls consist of three principal components:

(1) Definition of LLD's (2) Administrative limits for dose associated with liquid effluent releases (3) Dose tracking.

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4.2.1 The historical mix of radionuclides released in Rancho Seco liquid effluent provides the basis for,estab-lishing LLD's. The preceding discussion has shown that the presence or absence of dose associated with these releases can be very adequately indicated by a combin-ation of Csl34, Cs137 and I131. The concept of setting the LLD's at a concentration equivalent of 50% of the Technical Specification values has also been developed.

The resulting LLD's for liquid effluent offsite dose tracking purposes are shown in Table'4.

4.2.2 The District recognizes that the specification of LLD's a,t relatively high fraction of 0.5 of Technical Specif-7

ment approval is required. prior to all releases.

'The information flow for dose tracking will be as follows:

(Fig. 10)

(1) Prerelease dose calculation (2) Postrelease dose calculation (3) Weekly review of release records and update of dose calculation (4) Monthly composite of all RHUT A & B releases (5) Monthly inclusion of composite results in the dose !

calculation 4 (6) Monthly update of the remaining operating margin (7) Quarterly review of all release data, recalcu- i lation of dose and update of operating margin J

(8) Semiannual effluent report preparation.

The doses when calculated are plotted on a cumulative chart and compared to a target vector. (Fig. 11) The relative position of the cumulative dose line to the target is used as' basis for decisions to perform opera-tions cuch as recycling and demineralizing.

The update of the operating margin is performed to account for results obtained from the composite sample.

The purpose has both positive and negative aspects. On l

the positive side, the remaining operating margin may l

I be increased because the actual concentrations of radionuclides in the composite are less than the average of the LLD's and observable concentrations obtained in the release analyses. (Fig. 12) The negative is that the composite analyses may result in 9

un3xpsetcdly high average concentrations with an accom-panying reduction in the remaining operating margin.

4.3. Radiological Effluent Information Management System A computer-based effluent'information system is.in the initial stages of installation. A schematic of this system is shown in Figure 13. As indicated, Radiation Protection, Chemistry,' Environmental-Programs, Land Health Physics Support will'be linked to a common database.- One signif-icant: advantage to this system is that the personnel who perform prerelease dose calculations will be.using the same model as those who perform the followup weekly and monthly calculations. Features such as data-in-expected-range prompting'are to be included. Personnel will be trained and the system will b2 carefully phased in and customized to

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Rancho Seco over a period of a few months. Initial produc- ]

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tion operation is scheduled for August 1, 1987.

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'The basis.for a surveillance Technical Specification for 10 CFR Part 50 Appendix I compliance has been presented. Table 5 summarizes this Technical Specification. This Technical Specification establishes a method for compliance which takestinto account the unique.limnology of the Rancho seco f site and the limitations of a well-equipped process analyt-ical laboratory. It provides reasonable assurance that the design guides in Appendix I will not be substantially exceeded-even under. nominally adverse. conditions. As backup and confirmation of the effluent control program, the District operates a comprehensive environmental monitoring program in the liquid effluent pathway.

id REFERENCES 6.1 Oesterling, R. G., Development of Target Dose' Values.for Appendix i I compliance, March 16, 1987 4

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LOWER LIMIT OF DETECTION SPECIFICATION:

TECHNICAL SPECIFICATION TABLES 4.21-1, 4.21-2-and 4.22-1 (1) The lower limit of detection (LLD) for a radionuclides presented in this is table the largest concentration, expressed in microcuries per milliliter, which is required to be detected in. order to achieve compliance with the applicable regulation, given stated operating conditions and calculation methodology. 4 I

(2)' The LLD of a radioanalysis system is.that value which ;

will indicate the presence or absence of radioactivity.

in a sample when the probability of a false. positive and of a false. negative determination is stated. The probabilities of false positive and false negative are taken as equal at 0.05. The equation for LLD in

.'microcuries per milliliter is given by the equation:

LLD.=-

f[2.7 + 3.29(Br)-0.5) 3.7E4(YEVT) where 2.7 = factor to correct for Poisson distribution at very low background count rates

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f = correction factor to account for systematic errors = 1.1 B = background (counts) ts r = 1 + g-- (ts$t)d tb = background count time (seconds) o t s = sample count time (seconds) 3.7E4 = disintegrations /second/ microcurie Y= yield of radiochemical process-E= counting efficiency (disinteg-rations / count) .

V= sample volume (milliliters)

[1 - exp(-At s)]exp(-At c)

T=

A where X = decay constant (seconds-1) t c = time from collection to start of counting 1

BASIS FOR TECHNICAL SPECIFICATION LLD The equation for lower limit of detection is taken from Currie l, with minor modifications, as follows:

1. The systematic error has been determined to be negligible with respect to random error. The "f" term in Currie's equations has been retained to account for " unknown" systematic errors.

2. The factor "exp(-Atc )" to correct for decay time from collection to analysis is added for completeness.

3. The correction factor (2.7) for very low counts is added.

4. The nu symbol'was replaced by "r".

5. Changes were made to express concentration in microcuries per milliliter and counting time in seconds.

Reference

1. Currie, L. A., Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental' Measurements, NUREG/CR-4007, Sept. 1984 1

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IDWER LIMIT OF DETECTION SPECIFICATION:

' TECHNICAL SPECIFICATION TABLE 4.26 '

(1) The lower limit of. detection (LLD) for a radionuclides presented in this.is table the largest concentration, expressed ir.'microcuries per milliliter, which is required to be detected in order to achieve compliance with.the applicable regulation, given stated operating conditions and calculation methodology.

l (2) The LLD of a.radioanalysis system'isLthat'value which; will indicate the presence or absence of radioactivity

.in a sample when the probability of a. false positive and of.a false negative. determination is stated. The probabilities of false. positive-and false negative-are taken as equal at 0.05. The equation-for LLD in microcuries per milliliter is given by the equation:

LLD =

f(2.7 + 3.29(Br)-0.5) 0.037(YEVT).

where 2.7 = factor to correct for Poisson distribution at very low

, g background count rates f = correction factor to account for l

-systematic errors = 1.1 B = background (counts) ts r=1+ g-- b (ts5t) b tb = background' count time (seconds) t s = sample count time (seconds) 0.037 = disintegrations /second/ picocurie' Y= yield of radiochemical process E= counting efficiency (disinteg- ,

rations / count)

V= sample volume (liters) or mass (kilograms)

[1 - exp(-At s)]exp(-At c)

T= I h

where X = decay constant (seconds-1)~ ,

t c = time from collection to !

start of counting l 3

BASIS FOR TECHNICAL SPECIFICATION LLD The equation for lower limit of detection is taken from Currie l, with minor modifications, as follows:

1. The systematic error has been determined to be negligible with respect to random error. The "f" term in Currie's equations has been retained to-account for " unknown" systematic errors.

2. The factor "exp(-At c)" to correct for decay time from collection to analysis is added for completeness. ,

3. The correction factor (2.7) for very low counts is added.

4. The nu symbcl was replaced by "r". .

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3. Changes were made to express mass in kilograms and ,

counting time in seconds.

Reference

1. Currie, L., A., Lower Limit of Detection: Definition and Elaboration of a Pronosed Position for Radiological Effluent and Environmental Measurements, NUREG/CR-4007, Sept. 1984

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i 1981-84 HISTORICAL MIX OF RADIONUCLIDES IN RANCHO SECO LIQUID EFFLUENTS l RCTIVITY. l NUCLIDE )

FRACTION Cs137 0.307 Co58 0.261 Cs134 0.162 GRMMR Co60 0.068 Mn54 0.032 EMITTERS La140 0.063 Rg110m 0.006 EXCLUDING Cs\136 0.004 i Nb95 0.504 IODINES Cr51 0.003 Fe59 0.002 Ba190 0.001 Na24 8E-4 Zr95 BE-9 CoS7 ,

1E-4 Sr89 2E-5 I131 0.125 IODINES I133 0.015 TABLE 1

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f DOSE' FRACTIONS FOR EACH NUCLIDE FOR THE 1981-84 MIX DOSE FRACTION FOR NUCLIDE TOTAL LIMITING NUCLIDE BODY ORGAN Cs137 0.525 0.595 Cs134 0.453 .0.381 Co60 0.011 0.006 Mn54 0.004 0.006 CoS8 0.004 0.004 GRHHR Cs136 0.003 0.001 Ag110m 1E-4 SE-9 EMITTERS Fe56 1E-4 1E-4 g Hb95 6E-6 0.006 Ba140 4E-6 7E-4 Sr89 2E-6 2E-5 IODINES ZrS5 1E-6 2E-5 CoS7 4E-7 6E-7 Lo140 SE-7 SE-5 Cr51 1E-7 2E-6 Na24 7E-8 2E-8 1131 SE-9 0.9996 10 DINES 1133 SE-7 4E-4 TABLE 2

O G T N S N 5 1 S D S I

N TA 9 8 7 O E E O I G 5 3 S O T D I

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R . . . .

U L S T I O O O O 1 B CT C L I R R UN R T NE F T OU I

A L E L 5 3 8 D F S RY 2 5 7 E A F O R R E TD 5 4 9 S DOO . . .

N O TB O O O D G N D I

I R T U )

O T Q S E

F I I N I

S ML D O

R E- I O N G E

T RI M D 7 9 .

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i ATTACHMENT 2 PROPOSED AMENDMENT NO. 155

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__ ____________________-___________-_-_____--___]

4 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 1

TABLE OF CONTENTS 1

Section Page 1.16 RESTRICTED AREA 1-7 1.17 SITE BOUNDARY 1-7 1.18 DOSE EQUIYALENT I-131 1-7 ,

i 1.19 MEMBER (S) 0F THE PUBLIC 1-7 155> 1.21 MAXIMUM HYP0THETICAL INDIVIDUAL 1-8 i

1.22 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL 1-8 1.23 LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM 1-8 1.24 VENTILATION EXHAUST TREATMENT SYSTEMS 1-9 1.25 PURGE - PURGING 1-9 1.26 VENTING 1-9 1.27 RADI0 ACTIVE EFFLUENT 1-9 4 \

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 jfd 2.1 SAFETY LIMITS, REACTOR CORE 2-1 l 2.2 SAFETY LIMITS, REACTOR SYSTEM PRESSURE 2-4 !

2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 2-5 ,

3.0' LIMITING CONDITIONS FOR OPERATION 3-1 3.1' REACTOR COOLANT SYSTEM 3-1 3.1.1 Operational Components 3-1 3.1.2 Pressurization, Heatup, and Cooldown Limitations 3-3 3.1.3 Minimum Conditions for Criticality. 3-6 3.1.4 Reactor Coolant System Activity 3-8 l 3.1.5 Chemistry 3-10 Proposed Amendment No. 155 iia j

i l

RANCHO SEC0 UNIT 1 )

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) l l

Section Page (

3.14.5 Fire Hose Stations 3-57 3.14.6 Fire Barrier Penetration Fire Seals 3-58 155>< 3.15 RADI0 ACTIVE LIQUID EFFLUENT MCNITORING INSTRUMENTATION 3-60 3.16 RADI0 ACTIVE GASE0US EFFLUENT MlNITORING INSTRUMENTATION 3-63 3-70 3.17 LIQUID EFFLUENTS f 3-70 3.17.1 Concentration f 3.17.2 Dose 3-71 3.17.3 Liquid Holdup Tanks 3-72 155>< 3.17.4 Liquid Effluent Radwaste Treatment 3-72a 3.18 GASEOUS EFFLUENTS 3-73 3.18.1 Dose Rate 3-73 ISS> 3.18.2 Dose-Noble Gases 3-74 3.18.3 Dose-Iodine-131, Iodine-133 Tritium, and Radioactive Materials in Particulate Form 3-75 3.18.4 Gaseous Radwaste Treatment 3-78 3.18.5 Gas Storage Tanks 3-79 3.19 DELETED

< 3.20 DELETED 3.21 SOLID RADI0 ACTIVE WASTES 3-80 3.22 RADIOLOGICAL ENVIRONMENTAL MONITORING 3-81 3.23 LAND USE CENSUS 3-87 3 24 EXPLOSIVE GAS MIXTURE 3-89 3.25 FUEL CYCLE DOSE 3-90 3.26 INTERLABORATORY COMPARISON PROGRAM 3-92 Proposed Amendment No.155 iv

. RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section Page 4.14 SHOCK SUPPRESSORS (SNUBBERS) 4-47 4.15 RADI0 ACTIVE MATERIALS SOURCES 4-48 4.16 Reserved 4-49 4.17 STEAM GENERATORS 4-51 4.17.1 Steam Generator Sample Selection and Inspection 4-51 4.17.2 Steam Generator Tube Sample Selection and Inspection 4-51 4.17.3 Inspection Frequencies 4-52 4.17.4 Acceptance Criteria 4-53 4.17.5 Reports 4-54 4.17.6 OTSG Auxiliary Feedwater Header Surveillance 4-54 4.17.7 4-55 Inspectign Acceptance Criteria and Corrective Actions 4.17.8 Reports 4-55 g

i

$ 4-58 1 4.18 FIRE SUPPRESSION SYSTEM SURVEILLANCE 155>< 4.19 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 4-63 4.20 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4-65 4.21 LIQUID EFFLUENTS 4-69 l l

4.21.1 Concentration 4-69 155>< 4.21.2 Doses 4-72 ,

l 4.21.3 Liquid Holdup Tanks 4-73 j 155> 4.21.4 Liquid Effluent Radwaste Treatment 4-73a Proposed Amendment No. 155 vi i

l l

+

i RANCH 0 SECO UNIT 1 TECHNICAL: SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section Page 155> 4.22 GASE0US EFFLUENTS 4-74 4.22.1 Dose Rate. 4-74 4.22.2 Dose-Noble G&ses 4-77 4.22.3- Dose-Iodine-131, Iodine-133, Tritium, and Radioactive -

Materials in Particulate Form 4-79 4.22.4 Gaseous Radwaste Treatment 4-79

.4.22.5 Gas Storage Tanks ~ 4-80 4.23 Deleted 4-79 4.24 Deleted 4- 80 4 4.25- SOLID RADI0 ACTIVE WASTES 4-81 4.26 RADIOLOGICAL ENVIRONMENTAL MONITORING 4-83 1

4.27

\

LAND USE CENSUS 4-86 6- 4.28 EXPLOSIVE GAS MIXTURE 4-87 4.29 FUEL CYCLE DOSE 4-89 4.30 INTERLABORATORY COMPARIS0N PROGRAR SURVEILLANCE REQUIREMENT 4-90 4.31 NUCLEAR SERVICE ELECTRICAL BUILDING EMERGENCY HEATING 4-91 VENTILATION AND AIR CONDITIONING 5.0 DESIGN FEATURES 5-1 5.1 SITE 5-1 155> 5.1.1 Exclusion Area 5-1 5.1.2 Low Population Zone 5-1 5.1.3 Site Boundary for Gaseous Effluents 5-1 5.1.4 Site Boundary for Liquid Effluents 5-1 Proposed Amendment No. 155 vii

(

155>

. RANCHO SECO UNIT 1 4 TECHNICAL. SPECIFICATIONS .

. TABLE OF CONTENTS (Continued) i Section Page- l l

!5.2 CONTAINMENT 5-2 -

l i 5.2.1 Reactor Buf1 ding 5 1 5.2.2 Reactor Building Isolation System 5-3 l-5.3 REACTOR 5-4 i

5.3.1 Reacto'r Core 5 4.

5.3.2 Reactor Coolant System 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 New Fuel Inspection and Temporary Storage Rack 5-6' 5.4.2 New and Spent Fuel Storage Racks and Failed 5-6 Fuel Storage Container Rack 5.4.3 New and Spent Fuel Temporary Storage 5-6 5.4.4 Spent Fuel' Pool and Storage Rack Design 5-6 L

i j

Proposed Amendment No. 155 viia 4

3 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)'

Section Page 6 ADMINISTRATIVE CONTROLS 6-l' 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.3 FACILITY STAFF QUALIFICATIONS 6-3 l 6-3 6.4' TRAINING 6.5 REVIEW AND AUDIT 6-3 6.5.1 Plant Review Committee 6-3 6.5.2 Management Safety Review Committee 6-6 138> 6.5.3 Technical Review and Control 6-8 4 6.5.4 -Audits 6-9 6.6 REPORTABLE OCCURRENCE ACTION 6-10 s

6.7 SAFETY LIMIT VIOLATION 6-11 0! -

6.8 PROCEDURES 6-11 6.9 REPORTING REQUIREMENTS 6-12 6.10 RECORD RETENTION 6-13 6.11 RADIATION PROTECTION PROGRAM 6-14 6.12 RESPIRATORY PROTECTION PROGRAM - Deleted 6.13 HIGH RADIATION AREA 6-15 6.14 ENVIRONMENTAL QUALIFICATION 6-16 6.15 PROCESS CONTROL PROGRAM (PCP) 6-17 155> 6.16 0FFSITE DOSE CALCULATION AND RADIOLOGICAL MONITORING 4 PROGRAM MANUALS 6-18 (

6.17 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS l LLIQUlU, GAbtOUS, ANU SOLIUJ 6-19 j 6.18 POSTACCIDENT SAMPLING 6-22 Proposed Amendment No. 138, Rev. 2 J Proposed Amendment No. 155 .' .

viii

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES Section Page 1-10 155>< 1.9-1 FREQUENCY NOTATION 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 3-27 3.5.5-1 ACCIDENT MONITORING INSTRUMENTATION OPERABILITY REQUIREMENTS 3-38b 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3-40 3.7-1 V0LTAGE PROTECTION SYSTEM RELAY TRIP VALUES 3-41a 3.7-2 VOLTAGE PROTECTION SYSTEM LIMITING CONDITIONS 3-41b 3.12-1 SAFETY RELATED HYDRAULIC SNUBBERS 3-Sla-e 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55 3.14-1 INSIDE BUILDING FIRE HOSE STATIONS 3-57a 3.15-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3-61 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 3-64 3.22-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3-83 3.22-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS 3-86 IN ENVIRONMENTAL SAMPLES 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS 4-3 ,

4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY 4-8 ,

4.1-3 MINIMUM SAMPLING FREQUENCY 4-9 4.2-1 CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 4-12b 4.10-1 ENVIRONMENTAL RADIATION MONITORING PROGRAM. 4-42 4.10-2 OPERATIONAL ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-22a 1 Proposed Amendment No. 155 ix

4 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ]

LIST OF TABLES (Continued) :

Section Page 4.14-1 DESIGNATED SAFETY REi.ATED HYDRAULIC SNUBBERS FUNCTIONALLY 4-47d,e i TESTED ONLY AS REQUIRED BY THE. SNUBBER SEAL REPLACEMENT PROGRAM 4.17-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED 4-56 ,

1 DURING INSERVICE INSPECTION 4.17-2A STEAM GENERATOR TUBE INSPECTION 4-57 4.17-2B STEAM GENERATOR TUBE INSPECTION (SPECIAL LIMITED AREA) 4-57a 4.17-3 OTSG AUXILIARY FEEDWATER HEATER SURVEILLANCE 4-57b,c

-4.19-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 4-66 SURVEILLANCE REQUIREMENTS 4.21-1 .RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAN 4-70

.4.22-1 RADI0ACT'IVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM 4-75 4.26-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) 4-84 S:

4.28-1 EXPLOSIVE GAS MIXTURE INSTRUMENTATION SURVEILLANCE 4-88 REQUIREMENTS 138x 6.2-1 MINIMUM SHIFT CREW COMPOSITION 6-2 Proposed Amendment No.138, Rev. 2 Proposed Amendment No. 155 X

, 1

-RANCHO SECO UNIT 1

'. TECHNICAL SPECIFICATIONS.

LIST OF FIGURES (Continued)

Figure 3.5.2-7 Core Imbalance vs. Power Level, O to 40 EFPD.

3.5.2-8 ' Core Imbalance vs. Power Level, after 30 EFPD i

-3.5.2-9 Core Imbalance vs. Power Level, after 300.EFPD with APSRs Withdrawn 1

3.5.2-10 Deleted -

.3.5.2-11 Deleted 3.5.2-12' ' Deleted 3 5.4-1 Incore Instrumentation Specification Axial Imbalance Indication 3.5.4-2 Incore Instrumentation Specification Radial Flux Tilt Indication.

3.5.4-3 Incore Instrumentation Specification 1

3.18-1 General Layout of Site 4.13-1 Main Steam Inservice Inspection

'4.13-2 Main Feehwater Inservice Inspection

g - 4.13-3 Main. Steam Dump Inservice Inspection 155> 5.1-1 Exclusion Area 5.1-2 Low Population Zone 5.1-3 Site Boundary for Gaseous Effluents

< 5.1-4 Site Boundary for Liquid Effluents 138> 6.2-1 SMUD Corporate Support of Nuclear Safety to Rancho Seco

< 6.2-2 Nuclear Organization Chart Proposed Amendment No. 138, Rev. 2 Proposed Amendment No. 155 xii i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.11 FIRE SUPRESSION SYSTEMS 1.11.1 The FIRE SUPRESSION WATER SYSTEM shall consist of water sources, pumps and distribution piping with associated sectionalizing control of isolation valves. Such valves include yard hydrant ,

valves and the first valve ahead of the water flow alarm device on !

each sprinkler header, hose standpipe or spray system riser which- i protect nuclear safety components.

1.11.2 The FIRE SUPRESSION CARBON DIOXIDE SYSTEM shall consist of a CO2 source and distribution piping with sectionalizing control valves which protect nuclear safety components.

1.12 STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of: j

a. A test schedule for n systems, subsystems, trains or '

designated components obtained by dividing the specified test interval into n equal subintervals. I

b. The testing of one system, subsystem, train or designated components during each subinterval.

155> 1.13 PROCESS CONTROL PROGRAM g PROCESS CONTROL PROGRAM (PCP) - The PROCESS CONTROL PROGRAM shall j contain the sampling, analysis, and formulation determination by '

which SOLIDIFICATION of radioactive wastes from liquid systems is assured.

1.14 SOLIDIFICATION SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed),

monolithic, immobilized solid with definite volume and shape, ,

bounded by a stable surface of distinct outline on all sides j

( f ree-standing) . J 4 (

i Proposed Amendment No. 155 1-6 1

0 l

________________g

RANCHO SECO UNIT 1-TECHNICAL SPECIFICATIONS Definitions 1.15 0FFSITE DOSE CALCULATION MANUAL (0DCM) 155> The' 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall be a manual containing the description of the methodology, algorithms and parameters to be used in the calculation of offsite doses

< resulting from the release of radioactive material in gaseous and liquid effluents and in the calculation of gaseous and liquid 155>< effluent monitoring instrumentation alann/ trip setpoints.

1.16 RESTRICTED AREA That portion of the site property, the access to which is controlled by security fencing, equipment and personnel.

1.17 SITE BOUNDARY 155>< Site Boundaries are defined by Figures 5.1-1 through 5.1-4.

1.18 DOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose 155> via the inhalation pathway as the quantity and isotopic mixture.of

- I-131, I t132, I-133, I-134 and I-135 actually'present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP Publication 30, " Limits for Intakes of g < Radionuclides by Workers," 1979.

1.19 MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all individuals who by l virtue of their occupational status have no formal association 1 with the plant. This category shall include non-employees of the i licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall not include non-employees such as vending machine servicemen or postmen who,' as part of ,

their formal job function, occasionally enter an area that is !

controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

l Proposed Amendment No.155 1-7 t

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1 i

l RANCHO SECO UNIT 1 I TECHNICAL SPECIFICATIONS Definitions 155> 1.21 MAXIMUM HYPOTHETICAL INDIVIDUAL The MAXIMUM HYP0THETICAL INDIVIDUAL is characterized as " maximum" with regard to food consumption, occupancy, or other usage or exposure pathway parameters in the vicinity of Rancho Seco that a would represent an individual or composite of individuals with habits greater than usually expected for the average of the population in general. No single indi/idual would be expected to .

be exposed to all the potential effluent exposure pathways at the :

" maximum" val ue.

The MAXIMUM HYPOTHETICAL INDIVIDUAL is a hypothetical receptor of radiological exposure (mrem) resulting from the discharge of radioactive effluent (curies). The methodology to convert curies into mrem is described in the ODCM. The purpose of the ODCM calculation is to compare the resultant effluent exposure or dose with the numerical guides for design objectives in 10CFR60, Appendix 1.

The MAXIMUM HYPOTHETICAL INDIVIDUAL concept is consistent with its use in the US NRC Regulatory Guide 1.109 and 10CFR50, Appendix I.

The MAXIMUM HYP0THETICAL INDIVIDUAL concept is NOT used to demonstrate compliance with 10CFR20.106.

1.22 RADIOLOG'1 CAL ENVIRONMENTAL MONITORING PROGRAM (REMP) MANUAL j; The RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) MANUAL shall be a manual containing the description of the Rancho Seco radiological environmental monitoring program. The REMP MANUAL shall contain a description of the environmental samples to be collected, the sample locations, sampling frequencies, and sample analysis criteria. l 1.23 LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM The LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM is the system designed and installed to reduce the quantity of ' radioactive materials in liquid effluents by collecting liquid effluent and providing processing for the purpose of reducing the total radioactivity prior to its release to the environment.

i Proposed Amendment No.155 1-8

155>

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.24 VENTILATION EXHAUST TREATMENT SYSTEM The VENTILATION EXHAUST TREATMENT SYSTEMS are s;" tems designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for I

the purpose of removing iodines or particulate from the gaseous l

exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas I effluents). Engineered Safety Feature (ESF) atmospheric cleanup i

systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEMS components.

L l

l 1.25 PURGE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a i

manner that replacement air or gas is required to purify the confinement.

1.26 VENTING YENTING is the controlled process of discharging air or gas from a l confinement to maintain temperature, pressure, humidity, concentration ~ or other operating condition, in such a manner that replacement air or gas is not provided or required during

}$ VENTING. Vent, used in system names, does not imply a VENTING process.

1.27 RADI0 ACTIVE EFFLUENT Effluent shall be designated as RADI0 ACTIVE EFFLUENT when the radiochemical analysis of an appropriate sample of the effluent results in the detection of radioactive material above the Lower Limits of Detection as defined in the OFFSITE DOSE CALCULATION MANUAL.

E Proposed Amendment No. 155 1-9

(

155>

. TABLE 1.9-1 FREQUENCY NOTATION NOTATION- FREQUENCY-S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

W- .

' At least once per 7 days.

M At-least once per 31 days.

Q At least once per 92 days.

SY- At least once per 184 days.

A At least once.per 12 months R 'At least once per 18 months.

BA At least once per 24 months S/U Prior to each reactor startup.

P Completed prior to each release.

NA Not applicable.

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-Proposed Amendment No. 155 1-10 4

. RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.4 STEAM AND POWER CONVERSION SYSTEM Applicability

. Applies to the operability of the turbine cycle during normal operation and for the removal of decay heat.

0bjective To specify minimum conditions of the turbine cycle equipment necessary to assure the required steam relief capacity during nonnal operation and the capability to remove decay heat from the reactor core.

Specification 152> 3.4.1 The reactor coolant system shall not be brought or remain above 280F with irradiated fuel in the pressure vessel unless the following conditions are met:

A. . Capability to remove decay heat by use of two steam generators as specified in 3.1.1.2.

B. One atmospheric dump valve per steam generator shall be operable.

C. A minfrom of 250,000 gallons of water shall be available in i-the condensate storage tank.

D. Two main steam system safety valves are operable per steam generator.

E. Both auxiliary feedwater trains (i.e., pumps and their flow paths) are operable.

F. Both trains of main feedwater isolation on each main feedwater line are operable.

G. Four independent backup instrument air bottle supply systems for ADVs and MFW, SFW, and AFW valves are operable.

With less than the above required components operable, be on decay '

< heat cooling within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Proposed Amendment No. 152

. 3-23 l

- - - - - - - - - - - _ . - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Q

RANCHO SECO UNIT 1 TECHNICAL SPEC 1FICATIONS Limiting Conditions for Operation 155>< 3.15 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.15-1 shall be OPERABLE with their alarm / trip setpoints set to ensure 155> that the limits of Specification 3.17.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the -

< methodology contained in the OFFSITE DOSE CALCULATION MANUAL (0DCM).

Applicability During the release of radioactive effluents via the pathways identified in Table 3.15-1.

Action a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of Specification 3.17.1 are met, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take 155> the ACTION shown in Table 3.15-1. Exert best efforts to return the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Efflyent Release Report pursuant to Specification 6.9.2.3 why 4 the inoperability was not corrected in a timely manner.

g Bases 155>< During normal operations, radioactive contaminated water from primary system leaks and drains is processed in a liquid radwaste system and recycled into the Reactor Coolant Makeup System or otherwise, reused in the controlled areas 155> of the plant. Secondary system water is normally released from the plant.

The secondary system water, if it contains radioactive material is released

< through the ' A' and 'B' Regenerant Hold-Up Tanks (RHUTs). During periods of primary to secondary leakage, or when the sumps are contaminated, administrative controls require the turbine building sumps liq'u id effluent to 155>< be diverted to the ' A' and 'B' Regenerant Hold-Up Tanks.

155> Demineralized reactor coolant can be transferred from the demineralized Reactor Coolant Storage Tank (DRCST) to the ' A' and 'B' Regenerant Holdup Tanks for sampling, processing, and eventual discharge offsite as required by

< operational constraints.

Under normal conditions, the once through steam generators have no blow down. If a blow down is required during periods of priinary to secondary leakage, all water will be retained and processed in the radwaste system or diverted to the 'A' and 'B' Regenerant Hold-Up Tanks.

Proposed Amendment No.155 3-60

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS.

Limiting Conditions for Operation -

3.15 (Continued) 155> Bases ~ (Continued)

Radioactive liquid effluent monitoring instrumentation is provided to monitor and control,-as applicable, the releases of radioactive materials in liquid effluents during actual: or potential releases of radioactive liquid effluents. The alarm / trip setpoints for these instruments shall be calculated L

l: in accordance with the methodology contained in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.106.

The OPERABILITY and use of this instrumentation is consistent with the

, requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

\

4 Proposed Amendment No. 155 3-60a

.____________________ _ - -_-______ - ____________ _ _ m

RANCP.0 SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.15-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable Action

1. Gross Radioactivity Monitors Providing Automatic Termination of Release 155> a. Retention Basin 1 With the monitor inoperable, Effluent Discharge effluent releases may be resumed Monitor provided that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Specification 4.21.1.

2. At least two technically

\ qualified members of the Facility Staff independently n!

verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this pathway.

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l Proposed Amendment No. 155 3-61 ,

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RANCHO SECO UNIT 1

. TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.15-1 (Continued)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable Action 155> 2. Flow Measurement Devices

a. Regenerant Hold-Up 1 With the flow measurement device Tank Discharge Line inoperable, releases to the retention Total Flow' basins may continue provided the total flow can be determined by a tank level device or pump performance curves.

< b. Waste Water Flow Rate 1 With the flow rate measurement device inoperable, effluent releases via this pathway may continue provided the flow rate is estimated 155> at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during

\ retention basin releases.

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Proposed Amendment No. 155 3-62

)

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for. Operation 3.16- RADI0 ACTIVE GASEOUS EFFLUENT NONITORING INSTRUMENTATION The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.16-1 shall be OPERABLE with their alarm / trip setpoints set to ensure 155> that the limits of Specification 3.18.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the

< methodology contained in the ODCM.

Applicability During release via the pathways identified in Table 3.16-1.

Action a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value 155> which will ensure that the limits of Specification 3.18.1 are met, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably

< conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the 155> ACTION shown in Table 3.16-1. Exert best efforts to return the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Efflubnt Release Report pursuant to Specification 6.9.2.3 why 4 the inoperability was not corrected in a timely manner.

Bases 155> The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases radioactive of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated

< in accordance with the methodology contained in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.106.

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

155> The Auxiliary Building Stack is the effluent release point for the Waste Gas System and the Auxiliary Building Stack Noble Gas Activity monitor will ,

perform the necessary Waste Gas System release termination. The monitor alarms and terminates a Waste Gas Decay Tank release automatically

< if the activity exceeds the setpoint limits.

Proposed Amendment No. 155 3-63 i

RANCHO SECO UNIT 1 TECHN! CAL SPECIFICAT10NS 3.16 (Continued Limiting Conditions for Operation Bases (Continued) 155> The condenser air ejector exhaust has an individual noble gas monitor. This system exhausts into the Auxiliary Building ventilation system. Therefore,

. the Auxiliary Building Stack is the effluent release point and will alarm upon release of environmentally significant radioactive gases.

Fuel Storage Building exhaust is directed to the Auxiliary Building stack where the exhaust will be filtered and monitored for any activity prior to being released to the atmosphere, i

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Proposed Amendment No. 155 3-63a l

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I RANCHO SECO UNIT 1 .

TECHNICAL SPECIFICATIONS i

Limiting Conditions for Operation I

Table 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable Action

1. Reactor Building Purge i Vent

a. Noble Gas Activity 1 With the monitor channel alarm / trip Monitor providing setpoint less conservative alarm and automatic than required by Specification 155x termination of 3.18.1, immediately suspend the release. release or declare the channel inoperable. j With the monitor inoperable, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed in

\ accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. Iodine Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. Particulate Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Proposed Amendment No. 155 3-64

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.16-1 (Continued)

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable Action

2. Auxiliary Building Stack 155> a. Noble Gas Activity 1 With the monitor inoperable, Monitor providing effluent releases via this pathway alarm may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed in accordance with Table 4.22-1 within

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

b. Iodine Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and

\ these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. Particulate Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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Proposed Amendment No. 155 3-66 4

I

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RANCHO SECO UNIT 1-TECHNICAL SPECIFICATIONS Limiting Conditions for Operation i Table 3.16-1 (Continued)

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUHF3TATION Minimum Number of Channels Instrument Operable Action

2. ' Auxiliary Building

' Stack (continued)

d. System Effluent Flow 1 With the flow rate device Rate Device inoperable, effluent releases via this pathway may continue provided the flow rate used is the maximum design flow rate.

e. Sampler Flow Rate 1 With the flow rate device inoperable, '

Measuring Devices effluent releases via this pathway may continue provided the flow rate is estimated and recorded at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

155> f. Waste Gas , 1 With the monitor channel alarm /

l System (Auxiliary trip setpoint less conservative Building Stack than required by Specification g < Monitor) 3.18.1, immediately suspend the release or declare the channel inoperable.

With the monitor inoperable, the i

contents of the tank (s) may be released to the environment provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, and

b. At least two technically qualified members of the, Facil.ity Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.

Proposed Amendment No. 155 3-67

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS .

Limiting Conditions for Operation Table 3.16-1 (Continued)

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable Action 155> 3. Auxiliary Building

< Grade Level Vent

a. Noble Gas Activity 1 With the monitor channel alarm /

Monitor trip setpoint less conservative than 155>< required by Specification 3.18.1, immediately suspend the release or declare the channel inoperable.

With the monitor inoperable, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed in accordance with Table 4.22-1 within

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

b. Iodine Sampler 1 With the collection device

inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

155><

Proposed Amendment No. 155 3-68

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.16-1 (continued)

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable Action 155> 3. Auxiliary Building

< Grade Level Vent (continued)

c. Particulate Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue provided continuous samples are taken and these samples are analyzed in accordance with Table 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

d. System Effluent Flow 1 With the flow rate device Rate Device inoperable, effluent releases may continue provided the flow rate used is the maximum design flow rate.

\

e. Sampler Flow Rate 1 With the flow rate device g*

Measurement Device inoperable, effluent releases via this pathway may continue provided the flow rate is estimated and recorded at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

155><

Proposed Amendment No. 155 3-69 1

- - - - __ _ _ __ _ ___ _ __ _ )

i 155>

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation

)

3.17 LIQUID EFFLUENTS 3.17.1 Concentration 155> The concentration of radioactive material released in liquid

< effluents at any time beyond the Site Boundary For Liquid Effluents .{

(see Figure 5.1-4) shall be limited to the concentrations specified j in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrajned noble gases, the concentration shall be limited to 2x10- uCi/ml total activity.

Applicability . At-all times Action With the concentration of radioactive material released from the site.

155> exceeding Specification 3.17.1, immediately restore concentration within the-

< specification limits and report the event in the next Semiannual Radioactive Eff1 Lent Release Report' pursuant to Specification 6.9.2.3.

Bayes This Specification is provided to ensure that.t. he concentration of radioactive materials released in liquid waste effluents from the site to areas beyond the 155> Sitt Boundary For Liquid Effluent (see Figure 5.1-4) will be less than the y concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column

2. Dis limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in

< exposuras within the limits of 10 CFR Part 20.106 to MEMBER (S) 0F THE PUBLIC.

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

Proposed Amendment No. 155 3-70

RANCHO SECO UN1T 1 ;

TECHNICAL SPECIF1 CATIONS Limiting Conditions for Operation 3.17.2 Dose 155>< The dose or dose commitment to a MAXIMUM HYPOTHETICAL INDIVIDUAL from radioactive materials in liquid effluents released beyond the I 155>< Site Boundary For Liquid Effluents (see Figure 5.1.4) shall be limited to:

1 155> a. Less than or equal to 1.5 mrem to the total body and to less than or equal to 5.0 mrem to any organ during any calendar quarter; and, (

b. Less than or equal to 3 mrem to the total body and to less

< than or equal to 10 mrem to any organ during any calendar year.

Applicability At all times Action a. With the calculated dose or dose commitment from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 155>< 30 days a Special Report pursuant to Specification 6.9.5. This Report will identify the cause(s) for exceeding the limit (s) and define the corrective actions to be taken to reduce the 155>< releases of radioactive material in liquid effluents and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

Bases \

This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II. A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as 155>< low as is reasonably achievable." The dose calculation methodology in the ODCM implement the requirements in Section III.A of Appendix I that confomance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual i through appropriate pathways is unlikely to be substantially underestimated. l The equations specified in the ODCM for calculating the doses due to the i I

actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routir.a Releases of Reactor Effluents ,

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

155> Purpose of Implementing Appendix I," April 1977. There is reasonable assurance that the operation of the facility will not result in radionuclides concentrations in finished drinking water that are in excess of the 4 requirements of 40 CFR 141.

Proposed Amendment No.155 3-71 4

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1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation l 3.17.3 Liquid Holdup Tanks The quantity of radioactive material contained in each of the following tanks shall be limited to less than or equal to 10 Curies, excluding tritium and dissolved or entrained noble gases:

i 155> a. "A" and "B" Regenerant Holdup Tanks

b. Borated Water Storage Tank

c. Demineralized Reactor Coolant Storage Tank

d. Miscellaneous Water Holdup Tank j

< e. Outside Temporary Tanks Applicability At all times Action With the quantity of radioactive material in any of the listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to

'this condition in the next Serafonnual R,adioactive Effluent Release Report.

Bases 155> The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System or the LIQUID EFFLUENT RADWASTE TREATMENT

< SYSTEM.

Restricting the quantity of radioactive material contained in the specified outdoor tanks provides assurance that in the event of an 155> uncontrolled release of the tank's contents, the resulting

< concentration at the nearest portable. water supply and the nearest surface water supply in an unrestricted area would be less than the limits of 10 CFR 20, Appendix B, Table II, Column 2. The limit 155>< applies to each tank individually.

Proposed Amendment No. 155 3-72 8

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1 RANCHO SECO UNIT 1 (

TECHNICAL SPECIFICATIONS Limiting Conditions for Operation {

t 155> 3.17.4

Liquid Effluent Radwaste Treatment j I

The LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM shall be OPERABLE.

The appropriate portions of the system shall be used to reduce the quantity of radioactive materials in liquid effluents prior to their ,

di.scharge to ensure that projected doses due to the liquid effluent beyond the Site Boundary For Liquid Effluents (see Figure 5.1-4) will not exceed the requirements of Specification 3.17.2.

Applicability At all times.

Action

a. With the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM inoperable for more than 31 days or with radioactive liquid waste being df scharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.5 a Special Report which includes the following information:

1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,

~

2. Action (s) taken to restore the inoperable equipment to OPERABLE ctatus, and 3

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

Bases The OPERABILITY of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM were specified as the dose design objectives set forth in Section II.A of Appendix I,10 CFR Part 50, for liquid effluents. ,

The installation of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM is not complete. This specification will become effective when the system is

< declared operable.

Proposed Amendment No. 155 3-72a

RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS l Limiting Conditions for Operation j

3.18 GASEOUS EFFLUENTS ~

l 3.18.1 Dose Rate The dose rate due to radioactive materials released in gaseous 155> effluents from the site to areas at or beyond the Exclusion Area

< (see Figure 5.1-1) shall be limited to the following values:

155> a. The dose rate limit for noble gases shall be less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin; and,

b. The dose rate limit for Iodine-131, Iodine-133, tritium, and for all radioactive materials in particulate form with half lives greater than 8 days shall be less than or equal to 1500

< mrem /yr to any organ.

Applicability At all times Action 155> With the dose rate (s) cucceding the above limits, immediately restore the release rate to within the limit (s) given in Specification 3.18.1 and report the event in the next Semiannual Radioactive Ef, fluent Release Report pursuant

< to Specification 6.9.\2.3.

m Bases This specification is provided to ensure that the dose rate from gaseous 155> effluents at any time at the Exclusion Area Boundary (Figure 5.1-1) 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 concentration of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106 (b)(1)). For individuals who may at times be within tile Exclusion Area Boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates 4 above background to an individual at or beyond the Exclusion Area Boundary to less than or equal to 500 mrem /yr to the total body or to less than or equal to 3000 mrem /yr to the skin. These release rate limits also restrict, at all 155> times, the corresponding thyroid dose rate above background to an infant via 4 the inhalation pathway to less than or equal to 1500 mrem /yr.

Proposed Amendment No. 155 3-73

RANCHO SECO UNIT 1- ,

TECHNICAL SPECIFICATIONS I Limiting Conditions for Operation !

l Dose-Noble Gases j 155> 3.18.2 ).

The air dose due to noble gases released in gaseous effluents to areas at or beyond the Site Boundary For Gaseous Effluents (see Figure 5.1-3) shall be limited to the following:

a. During any calendar quarter, to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation; and,

b. During any calendar year, to less than or equal to 10 mrad for gamma radiation and to less than or equal to 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, prepare and submit to the Commission within 30 days a Special Report pursuant to Specification 6.9.5. This Report will identify the cause(s) for exceeding the limit (s) and define the corrective action (s) to be 155>< taken to rgduce the release of radioactive noble gases in gaseous effluents dnd the proposed corrective action (s) to be taken to assure that subsequent releases will be in compliance with the above g limits.

Bases This specification is provided to implement the requirements of Sections II.B.

III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". The Surve11' lance Requirements impigient 155> the requirements in Section III. A of Appendix I that conformance with the ' i guides of Appendix I 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 underesti. mated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109.

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpcse of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision 1. October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. The ODCM equations provided for determining that the air doses at the Site Boundary for Gaseous Effluents (Figure 5.1-3) are based upon the historical average atmospheric

< conditions.

Proposed Amendment No. 155 3-74

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 155> 3.18.3 Dose-Iodine-131, Iodine-133, Tritium and Radioactive Materials in j Particulate Form.

The dose or dose commitment to a MAXIMUM HYPOTHETICAL INDIVIDUAL from Jodine-131, Iodine-133, tritium, and radioactive materials in particulate form with half-lives greater than eight days in gaseous effluents released to areas at or beyond the Site Boundary for Gaseous Effluents (see Figure 5.1-3) shall be limited to the.

following:

a. During any calendar quarter to less than or equal to 7.5 mrem

, to any organ; and,

b. During any calendar year, to less than or equal to 15 mrem to 4 any organ.

Applicability At all times Action 1 With the calculated dose or dose commitment from the release of 155> Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form with half-lives greater than eight days in gaseous effluents exceeding any of the above limits, prepare and submit to the Commiss' ion within 30 days a Special Report pursuant to Specification 6.9.5. This Report will identify the cause(s) for i exceeding the limit and defines the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent release will be in compliance with

< the above annual limits.

Bases 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 1. The ACTION statements provide the required operating flexibilityIand at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as .

low as is reasonably achievable." The ODCM calculational methods specified in ;

the surveillance requirements implement the requirements in Section III.A of -

Appendix I that conformance with the guides of Appendix I be shown by j calculational procedures based on models and data, such that the actual {

exposure of an individual through appropriate pathways is unlikely to be !

substantially underestimated.

155><

l Proposed Amendment No.155 i 3-75 , j 1

-.______________________________________________________________________________________J

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.18.3 (continued)

Bases (continued) 155> The ODCM calculatiorjal methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. These equations also provide for estimating doses based upon the

< historical average atmospheric conditions.

155> The release rate specifications for radioiodines and radioactive materials in particulate form are dependent on the existing radionuclides pathways to man in

< areas at or beyond the Site Boundary for Gaseous Effluents (Figure 5.1-3).

The pathways which were examined in the development of these calculations are: (1) individual inhalation of airborne radionuclides, (21 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.

\

Proposed Amendment No.155 3-76 e

j

' RANCHO SECO UNIT 1 '

TECHNICAL SPECIFICATIONS.

LIMITING CONDITIONS FOR OPERATION PA 155>

This page has been intentionally left blank.

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Proposed Amendment No. 155 3-77

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 155> 3.18.4 Gaseous Radwaste Treatment ,

The Waste Gas System and the VENTILATION EXHAUST TREATMENT SYSTEM '

shall be OPERABLE. The appropriate portions of these systems shall ;

be used to reduce radioactive materials in gaseous waste prior to i their discharge such that projected gaseous effluent to areas at and beyond the Site Boundary for Gaseous Effluents (see Figure 5.1-3)

< are within the requirements of Specifications 3.18.2 and 3.18.3.

155>< App 1fcability At all times.

Action

a. With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 155>< days, a Special Report pursuant to Specification 6.9.5 which includes the following information:

1. Explanation of why gaseous radwaste was being discharged without treatment, identification of the equipment or ,

subsystems not OPERABLE and the reason for inoperability.

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

\

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

Bases 155>< The OPERABILITY of the Waste Gas System and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The 155> requirement that the appropriate portions of these systems be used, when

< specified, provides reasonable assurance that the releases of ra'dioactive 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 Appendixi A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were 155>< specified as the dose design objectives set forth in . Sections II.B and II.C of i Appendix I,10 CFR Part 50, for gaseous effluents. ;

i Proposed Amendment No. 155 3-78

)

I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 155>< 3.18.5 Gas Storage Tanks The quantity of radioactivity contained in each waste gas decay tank 155>< shall be limited to less than or equal to 135,000 curies of noble gases (considered as Xe-133).

Applicability At all times Action 155> a. With the quantity of radioactive material in any waste gas decay tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release

< Report, pursuant to Specification 6.9.2.3.

Bases Restricting the quantity of radioactivity contained in each waste gas decay 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 155>< exclusion area boundary (See Figure 5.1-1) will not exceed 500 mrem. This is

nnsistent with Standard Review Plan 15.7.1, " Waste Gas System Failure."

Potentw1 atmospherth releases from a waste gas decay tank are evaluated assuming 631gn coolant activities (see page 140-25 Vol. VI FSAR). Based on i; primary coolant activity as shown in Table 14D-7, the decay tank is assumed to hold the activity associated with the off-gas from one' reactor coolant system degassing with no credit taken for decay.

Calculation of the limiting dicay tank activity based on the coolant activity limit of Technical Specification 3.1.4 yields a maximum decay tank inventory of 98,414 C1 (Ref. FSAR Table 14D-23). In order for the decay tank inventory to reach the limiting condition "or operation, coolant activity would have to exceed the Technical Specification 3.1.4 limit on coolant activity and this would require a reactor shutdown, thus preventing a further increase in gaseous activity.

Therefore, it is conservative to require that the online waste gas decay tank 155>< be sampled daily upon reaching the reactor coolant system limiting activity value (43/E) to ensure the 135,000 curies equivalent Xe-133 is not exceeded.

Once the coolant is below the limiting activity, there is no requirement to sample waste gas decay tanks except for discharging.

Proposed Amendment No.155 3-79

RANCHO SECO UNIT 1 ~

1 TECHNICAL SPECIFICATIONS l Limiting Conditions for Operation 3.21 SOLID RADI0 ACTIVE WASTES l

155> The solid radwaste systems shall be OPERABLE and used in accordance with a PROCESS CONTROL PROGRAM for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements of 10 CFR 20 and 10 CFR 71 prior to shipment of radioactive wastes from the

< site.

Applicability At all times Action 155>- a. With the provisions of the PROCESS CONTROL PROGRAM net satisfiea, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

b. With the solid radwaste system inoperable for more than 31 days, prepare and submit to the commission within 30 days pursuant to Specification 6.9.5 a Special Report which includes the following information:

1. Identification of the inoperable equipment or sub-systems and the reason for inoperability,

2. Actio'n(s) taken to restore the inoperable equipment to OPERABLE status,

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

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

Bases The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 e CFR 50.36a and General D.esign Criterion 60 of Appendix A to 10 CFR 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may 155>< include, but are not limited to waste type, waste pH, waste / liquid /

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

Proposed Amendment No. 155 3-80 ,

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.22 RADIOLOGICAL ENVIRONMENTAL MONITORING

- I 155>< The Radiological Environmental Monitoring Program shall be conducted as specified in Table 3.22-1.

Applicability At all times Action 155>< a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 3.22-1, prepare and submit to the 155> Commission, in the Annual Radiological Environmental Operating

< Report required by Specification 6.9.2.2, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. (Deviations are permitted from the required sampling schedule if specimens are unobtainable due to 155>< hazardous conditions, or seasonal unavailability.)

b. With the level of radioactivity in an environmental sampling medium 155>< exceeding the reporting level of Table 3.22-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days after the level of radioactivity has been determined, a Special 155>< Report pursuant to Specification 6.9.5 which includes an evaluation of any release conditions, environmental factors or other aspects which caused the reporting limits to be exceeded. This report will define cor'rective actions to reduce emissions such that potential 155> exposures will meet Specification 3.25. When more than one of the y

radionuclides in Table 3.22-2 are detected in the sampling mediuin,

< this report shall be submitted if:

Concentration (1) + Concentration (2) + . . . >1.0 reporting level (1) reporting level m 155> When radionuclides other than those in Table 3.22-2 are detected and are the result of plant ef fluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.17.2, 3.18.2, and

< 3.18.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

155>< c. With milk or fresh leafy vegetation samples unavailable from any of the sample locations required by Table 3.22-1, prepare and submit to 155> the Commission within 30 days a Special Report pursuant to

< Specification 6.9.5 which identifies the cause of the unavailability of samples and identifies locations for obtaining replacement samples. The locations from which samples were unavailable may then be deleted from Table 3.22-1 provided the locations from which the 155> replacement samples were obtained are added to the Radiological

< Environmental Monitoring Program as replacement locations, if available.

Proposed Amendment No. 155 3-81

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS

- Limiting Conditions for Operation 3.22 (continued)

Bases 155>< The Radiological Environmental 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 155> station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and ODCM modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmer.tal Monitoring, Revision 1, November 1979. The specified monitoring program is in effect at this time. Program changes may be initiated based on operational experience, and changes in regional population or agricultural practices. The sample locations have been listed in the RADIOLOGICAL ENVIRONMENTAL MONITORING

< PROGRAM (REMP) MANUAL to retain flexibility for making changes as needed.

155> The detection capabilities required in Table 4.26-1 are state-of-the-art for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirement of 40 CFR 141.

< \ ,

iib l

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1 Proposed Amendment No. 155 3-82

- J

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for-0peration Table 3.22-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Sampling and 4 Exposure Pathwe Number of Collection Type and Frequency and/or Sample Samples

Fre'quency of Analysis j

1. AIRBORNE

'A. Radiciodine 8 Continuous oper-~ Radioiodine canis-and Parti- ation of sampler ter. Analyze at 155>< culates with sample least once weekly collection as for I-131.

required by dust loading but at Particulate least once per sampler. Analyze week. for Gross Beta radioactivity at 155><

least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following filter change. Perform gamma isotopic analysis on each sample where gross

\ beta activity is greater than 10 times the yearly

} 155>

mean of control 4

samples for the same sample period.

Perform gamma iso-topic analysis on composite (by i

location) for particulate filters sample at least once per quarter.

2. DIRECT Greater than 40 At least once Gamma dose. At RADIATION locations with 2 per quarter, least once per dosimeters at each quarter.

location.

155><

Sample locations are shown in the REMP MANUAL.

Proposed Amendment No. 155 3-83

.,4

<=

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ,

Limitir.g Conditions for Operation Table 3422-1 (Continued)

. RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Sampling and Exposure Pathwe Number of Collection Type and frequency and/or Sample Samples

Frequency of Analysis

3. WATERBORNE 155> a. Surface 1 Composite Gamma isotopic sample collected and tritium monthly ** analysis of each composite.

3 Grab sample Gamma isotopic and collected tritium analysis of monthly. each sample.

l b. Runoff 1 Grad enmple Gamma isotopic and collected tritium analysis of fortnightly. each sample,

c. Ground \2 At least once Gamma isotopic, I. per quarter. and tritium u analysis of each sample.

f l

l d. Mud and 2 At least once Gamma Isotopic l Silt semi-annually. analysis of

< One pint sample each sample.

of the top 3" .

of material 2 ft. from shoreline.

Sample locations are shown in the REMP MANUkL.

155>< ** Applicable when sampler is declared operational.

Proposed Amendment No. 155 3-84 j

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.22-1 (Continued) -

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Sampling and l Exposure Pathway Number of Collection Type and frequency and/or Sample Samples

Frequency of Analysis

4. INGESTION 155> a. Milk 4 At least weekly Gamma isotopic -

when animals analysis and are on pasture; I-131 analysis of L at least once each sample.

per month at other times.

b. Fish and 3 At least Gamma isotopic Inverte- quarterly. One analysis on edible brates sample of each portion of each species as sample.

listed in the REMP MANUAL.

I

c. Food \ 4 At time of har- Gamma isotopic i vest. One sam- analysis on pie of each of edible portion

< the several of each sample.

classes of food products as shown in the REMP MANUAL.

Sample locations are identified in the REMP MANUAL.

155><

Proposed Amendment No. 155 3-85 l

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5 1

RANCH 0=SECO UN!T 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.23 LAND USE CENSUS A land use census shall-be conducted annually and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden

of greater than 500 square feet producing fresh 155>< leafy vegetation in each of the 16 meteorological sectors within a distance of five miles.

155> The Land Use Census shall also include information relevant to the liquid effluent pathway and gaseous effluent pathway such that the 0FFSITE DOSE CALCULATION MANUAL (0DCM) and the RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL can be kept current witn the-

< existing environmental and societal uses surrounding Rancho Seco.

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 155>< being calculated in Specifications 4.21.2, and 4.22.3, identify the new locations in the next Annual Radiological- Environmental l Operating Report.

1

b. With a lan'd use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway)-

g 20 percent greater than at a location from which samples are l

currently being obtained in accordance with Specification 3.22, add l

the new location (s) to the Radiological Environmental Monitoring l 155> Program within 30 days or submit a Special Report to the Commission l- pursuant to Specification 6.9.5 that identifies the cause(s) for exceeding these requirements and the proposed corrective actions for

< precluding recurrence. The sampling location (s), excluding the l

control station location, having the lowest calculated dose or dose commitment (s) (via the same exposure pathway) may be deleted from this monitoring program af ter (October 31) of the year in which this

. land use census was conducted. Identify the new location (s) in the 155>< next Annual Radiological Environmental Operati.ng Report and also include in the report a revised figure (s) and table for the REHP manual reflecting the new location (s).

Broad leaf vegetation sampling may be performed at the site boundary in the 155>< direction sector with the highest D/Q in lieu of the garden census.

Proposed Amendment No. 155 3-87

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 3.23 (Continued)

Bases This specification is provided to ensure that changes in the use of 155> unrestricted areas are identified and that modifications to the Radiological

< Environmental Monito' ring Program and the 00CM 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 vegetation will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used: (1) that 20 percent of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage); and (2) a vegetation yield of 2 kg/ square meter.

155> In addition, by gathering information on the liquid effluent pathway and the gaseous effluent pathway, the census will ensure that proper radiological environmental monitoring and radioactive effluent controls are in place for 4 the adequate protection of the health and safety of the general public.

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$i Proposed Amendment No.155 3-88 '

4

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.25 FUEL CYCLE DOSE 155> The dose or dose commitment to any real MEMBER OF THE PUBLIC due to releases of radioactive material in gaseous and liquid effluents and to direct radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the . thyroid, which is

< limited to less than or equal to 75 mrem) over 12 consecutive months.

Applicability At all times Action 155> a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.17.2.a 3.17. 2.b , 3.18.2.a . 3.18. 2.b, 3.18. 3.a or 3.18.3.b, or exceeding the reporting levels of Table 3.22-2, calculations shall be made including direct radiation contributions (including outside storage '

tanks, etc.) to determine whether the above limits of Specification 3.25

, have been exceeded.

b. If the above limits have been exceeded, prepare and submit to the Commission within 30 days, a Special Report pursuant to Specification 6.9.5 that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule fo This Special Report,\r achieving as defined in 10conformance with the CFR Part 20.405(c), shallabove includelimits.

an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE ,

W P.UBLIC from uranium fuel cycle sources, including all effluent pathways !

~

and direct radiation, over 12 consecutive months that includes, the release (s) covered by 'this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations.

c. If the estimated dose (s) exceed the above limits, and if the release condition resulting in the violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provision of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until

< staff action on the request is complete.

Bases l 155>

3s specification is provided to meet the dose limitations of 40 CFR 190 that nave been incorporated into 10 CFR 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the numerical guides for design objective doses of Appendix I or exceeds the reporting levels for the Radiological Environmental Monitoring Program. For the Rancho Seco site it is unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the plant remains within twice the numerical guides for design objectives of 10 CFR 50 Appendix I and if 4 direct radiation (outside storage tanks, etc.J is kept small. The Special Proposed Amendment No. 155 3-90 l

RANCHO SECO UN!T 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 1 3.25 (Continued) - ,

, Bases (Continued) 155> Report will describe a course of action which should result in the limitation -

of the dose to a HEMBER OF THE PUBLIC for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumeo ,

that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any MEMBER OF THE PUBLIC is evaluated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190 is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is 4 engaged in carrying out any operation which is part of the uranium fuel cycle.

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B 1

Proposed Amendment No. 155 3-91 4

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conoitions for Operation 1

3.26 INTERLABORATORY COMPARIS0N PROGRAM The contractor performing the analysis of radiological' environmental i 155>< monitoring samples for radioactive materials shall participate in an Interlaboratory Comparison Program approved by the Commission.

Applicability At all times I

Action With analyses not being performed as required above,. report the corrective actions taken to prevent a recurrence to the t.ommission 155> in the Annual Radiological Environmental Operating Report pursuant

< to Specification 6.9.2.2.

Bases ]

The requirement for participation in an Interlaboratory Comparison Program is- '

provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid for .

i the purposes of Section IV.B.2 of Appendix I to 10 CFR 50.

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Proposed Amendment No. 155 3-92 l

_ _ _ _ _ - _ _ _ _ - _ _ - _ l

l RANCHO SECO tatti 1 -

TECHNICAL SPECIFICATI0uS '

Survelliance Standards Table 4.1-1 (Continued)

. INSTRUMENT SURVEILLANCE REQU1RaetTS , ,

I Chamael Description Cheet Test Calibrate Renarts !

42. Reactor Building drain NA NA R accumulation taak level NA .~

43. Incore neutron detectors M(1) NA .

(1) Checkfunctiontag including functioning,o f cogster readout and/or recorder readout.

155, 44. a. Process radiation annitoring systen W Q R

, b. Area radiation

< uenitoring systen W M Q k NA R

c. Contatrument Ana Monitors

45. Emergency plant radiation (1) 84ttery chect M(1) NA R Instevnents M(1) NA R (1) Check functioning

46. Environmental air monitors

47. Strong motion accelerometer Q(1) NA R (1) Battery check

48. Auxiliary Feedwater Start Ctrtuit

a. Phase tabalance/Under%

Power RCP *Y.

\ S 'NA R b.LowMainj'eedwater , -

6 Pressure WA N R .

M NA R

49. Pressurizer Water Level -

50. Auxiliary Feedwater Flow M NA R Rate

51. Reactor Coolant System -

Subcooling Marsin Monitor M NA R 52 DMOV Power Position Indicator (Primary Detector) H HA R

53. EHOV Position Indicator (Backup Detector)

T/C or Acoustic M NA R i

54. EMOV Block Valve Position Indicator M NA R

55. Safety Valve Position Indicator (Primary Detector)

M NA R T/C 56, Safety Yalve Position Indicator (Backup Detector)

Acoustic M MA R Proposed Araendment 155

(. 7h

RANCHO SECO UNIT 1 j

TECHNICAL SPECIFICATIONS 4 I

Surveillance Standards l f

155>< 4.19 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION {

Surveillance Requirements 155> The maximum setpoint shall be determined in accordance with methodology as described in the Offsite Dose Calculation Manual (ODCM) and shall be recorded

< on the release permits.

Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the INSTRUMENT CHANNEL CHECK, SOURCE CHECK, INSTRUMENT CHANNEL CALIBRATION, AND CHANNEL TEST at the frequencies shown in Table 4.19-1.

Records shall be maintained in the Process Standards of all radioactive liquid effluent monitoring instrumentation alarm / trip setpoints. Maximum setpoints j and setpoint calculations shall be available for review to ensure that the 1 155>< limits of Specification 3.17.1 are met.

Bases 155> The radioactive liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in

< liquid effluents during actual or potential release of radioactive liquid effluents. The alarm / trip setpoints for these instruments shall be calculated 155>< in accordance with the methodology contained in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.106. '

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

Proposed Amendment No. 155 4-63 :

I

RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.19-l' RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Instrument Channel Source Channel Channel Instrument Check Check Calibration Test 1.- Gross Radioactivity Monitors Providing Alarm and Automatic isolation 155> a. Retention Basin D P RI2) Q(1)

Effluent Discharge Monitor

2. Flow Monitors

a. Regenerant Hold-up D(3) NA R Q Tank Discharge Line Total Flow

b. Waste Water Flcw D(3) NA R -Q Rate \

(

t Proposed Amendment No. 155 4-64 4

i i

RANCHO SECO UNIT 1-TECHNICAL SPECIFICATIONS Surveillance Standards 155> TABLE 4.19-1 (Continued)

TABLE NOTATION (1) The CHANNEL TEST shall also demonstrate that automatic isolation of Chis pathway and control room alarm annunciation occui's if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm / trip sett aint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate raode.

1 (2) The INSTRUMENT CHANNEL CALIBRATION shall be performed using one or more !

reference standards (3) The INSTRUMENT CHANNEL CHECK shall consist of verifying indication of flow during periods of release. INSTRUMENT CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.

D l

Proposed Amendment No. 155 4-64a

RANCHO SECO UNIT 1 1 TECHNICAL SPECIFICATIONS J Surveillance Standards 4.20 RADI0 ACTIVE GASE0US EFFLUENT NONITORING INSTRUMENTATION l Surveillance Requirements 155> The maximum setpoints 'shall be determined in accordance with methodology described in the 0FFSITE DOSE CALCULATION MANUAL (ODCM) and shall be

< recorded on release permits. ;

Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the INSTRUMENT CHANNEL CHECK, SOURCE 1 CHECK, INSTRUMENT CHANNEL CALIBRATION, AND CHANNEL TEST at the frequencies shown in Table 4.20-1.

Records shall be maintained in the Process Standards of all radioactive {

gaseous effluent monitoring instrumentation alarm / trip setpoints. Maximum setpoints and setpoint calculations shall be available for review to ensure 155x that the limits of Specification 3.18.1 are met.

Bases The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in 155> gaseous effluents during actual or potential releases of radioactive gaseous effluents. The alarm / trip setpoints for.these instruments'shall be calculated

< in accordance with the methodology contained in the ODCM to ensure that the alam/ trip will occuh prior to exceeding the limits of 10 CFR Part 20.106. l The OPERABILITY and use of this instrumentation is consistent with the p'

requirements and General Design' Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The flow rates in the Auxiliary Building Stack and Auxiliary Building Grade Level Vent are constant as they use single speed fans. The Reactor Building Purge Vent has three different flow rates, winter, summer and mini-purge, however administrative controls assure using the correct flow rate where applicable. The actual flow rate of the ventilation systems are periodically determined by surveillance procedures. The flow rate measurement devices are used only as flow indicating devices and not for actual measurement of flow ,

rate. Also, as these flow rate devices must be removed from the ventilation l system for the channel test, and in addition transported to the manufacturer i for calibration, the frequencies have been- set as shown in Table 4.20-1.

1 Proposed Amendment No. 155 l 4-65 )

{

--- _d

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.20-1 ,

RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Instrument Channel Source Channel Channel Instrument Check Check Calibration ' . Test

1. Reactor Building Purge Vent

a. Noble Gas 155>< Activity Monitor D MI4I R(3) g(1)

b. Iodine Sampler W NA NA NA

c. Particulate Sampier W- NA NA NA

d. System Effluent 155>< Flow Rate Device D NA R Q(6)

e. Sampler Mohitor Flow Rate g; 155>< Measurement Device D NA R- Q

2. Auxiliary Building Stack

a. Noble Gas 155>< Activity Moniter D(5) M R(3) g(7)

b. Iodine Sampler W NA NA NA

c. Particulate Sampler W NA NA NA

d. System Effluent 155>< Flow Rate Device D NA R Q(6)

e. Sampler Monitor Flow Rate 155>< Measurement Device D NA R Q 155><

Proposed Amendment No. 155 4-66

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.20 (Continued)

Instrument .

Instrument Channel Source' Channel Channel Instruments Check Check Calibration Test 155>< 3. Auxiliary Building Grade Level-Vent 155>< a. Noble Gas D M R(3) g(2)

Activity Monitor

b. -Iodine Sampler W NA NA NA

c. Particulate W NA .NA NA Sampler

.155>< d. System Effluent D NA R Q Flow Rate Device 155>< e. Sampler Monitor D NA R Q Flow Rate Measurement Device.

\

155>< ,

f Proposed Amendment No. 155 4-67

155>-

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards ,

Table 4.20-1 (Continued) {

TABLE NOTATION l l

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

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

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

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

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

2. Circuit failure.

}

3. Instrument indicates a downscale failure.

\

4. Instrument controls not set in operate mode.

E (3) The INSTRUMENT CHANNEL CALIBRATION shall be performed using one or more reference standards.

(4) A check shall be performed prior to each release.

(5) A check shall be performed prior to each release via a Waste Gas Decay Tank ( s) .

(6) To be performed when device is accessible and conditions do not pose a i personnel safety hazard (i.e., potential main steam safety actuation). '

(7) The CHANNEL TEST shall also demons'trate that the Waste Gas System automatically isolates and that control room annunciation occurs if any of the following conditions exist:,

I

1. Instrument indictes measured levels above the ' alarm / trip setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

Proposed Amendment No. 155 4-68

RANCHO SECO UNfT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.21 LIQUID EFFLUENTS 4.21.1 Concentration Surveillance Requirements l

The concentration of radioactive material at any time.in liquid effluents released from the site shall be continuously monitored in accordance with Table.3.15-1.

The liquid effluent continuous monitor having provisions for automatic termination of liquid releases, as listed in Table 3.15-1, shall be used to limit the concentration of radioactive material released at any time from the 155>< site to areas beyond the site boundary to the limits given in Specification 3.17.1.

The radioactivity content of each batch of liquid effluent to be discharged shall be determined prior to release by sampling ano analysis in accordance with Table 4.21-1. The results of pre-release analyses shall be used with the 155> calculational methods in the OFFSITE DOSE CALCULATION MANUAL (0DCM) to assure

< that the concentration at the point of release is limited to the limits of Specification 3.17.1.

155>< '

Bases \

q This Specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to areas beyond the site boundary for liquid effluent will be less than the concentration levels 155>< specified in 10 CFR Part 20, Appendix B, Table II, Column 2.

This limitation provides additional assurance that the levels of radioactive' materials in bodies of water outside the site will result in exposures within 155>< the limits of 10 CFR Part 20.106 to MEMBER (S) 0F THE PUBLIC. The concentration ifmit for dissolved or entrained . noble gases is based upon the assumption that Xe-135 is the controlling radioisotopes and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

155> There are no continuous releases of radioactive material in liquid effluents from the plant. All radioactive liquid effluent releases from the plant are hy batch method. '

Proposed Amendment No. 155 4-69

_ _ _ _ _ _ _ _ _ _ _ _ _ ___ a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Tab 1'e 4.21-1 RADI0 ACTIVE LIQUID UASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Analysis Type of Activity Lower Limit Liquid Release Frequency Frequency Analysis Of Detection Type (LLD)

(uCi/ml)(a)

Each Batch Mn-54, Fe-59, 5 x 10-7 A. Batch Waste lease Tanks b,d(Re-) Each Batch Co-58, Co-60 P P Zn-65, Ho-99, Cs-134, Cs-137 l Ce-141, and !

Ce-144 (c) 155> I-131 3 x 10-7 Dissolved and 1 x 10-5 Entrained Gases

\ (Gamma Emitters)

H-3 1 x 10-5

(

i Proposed Amendment No. 155 4-70

)

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 4 Surveillance Standards-TABLE 4.21-1 (Continued)-

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM i l

Table Notation 155> a. (1) The lower limit of detection (LLD) for a radionuclides presented in-this table is the largest concentration, expressed in microcuries per milliliter, which is required to be . detected, if present, in order to achieve compliance with the . limits of Specification 3.17.1 (10CFR20, Appendix B, Table II, Column 2).

(2) The LLD of a radioanalysis system is that value which will indicate the presence or absence of radioactivity in a sample when the probability of a false positive and of a false negative detennination is stated. The probabilities of false positive and false negative are taken as equal at 0.05. The equation for LLD in microcuries per milliliter is given by the equation:

LLD = f[2.7 + 3.29(Br)-0.5]

3.7E4(YEVi) where 2.7 = factor to correct for Poisson

\ distribution at very low background count rates.

f = correction factor to account for systematic 3:

errors = 1.1.

B = background (counts) r=1+

t s (t3<t)h b

tb = background count time iseconds) f t3 = sample count time (seconds) )

3.7E4 = disintegrations /second/ picocurie I

I Y = yield of radiochemical process E'= counting efficiency (disintegrations / count)

V = sample volume (liters) or mass (kilograms)

Proposed Amendment No. 155 4-71 1

i

l i RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS Surveillance Standards l

4.21.2 Doses i Dose Calculations 155> Cumulative dose contributions and cumulative dose projections associated with the release of liquid RADI0 ACTIVE EFFLUENTS from the site (see Figure 5.1-4) shall be determined in accordance with the sampling and analyses specified in :

Table 4.21-2 and the methodology described in the Offsite Dose Calculation Manual (0DCM) at the following frequencies:

a. Prior to the initiation of a release of liquid RADI0 ACTIVE EFFLUENT; and,

b. Weekly, based on gamma-emitter and tritium analyses of liquid RADI0 ACTIVE EFFLUENT released during the previous seven days; and,

c. Monthly, based on gamma-emitter and tritium analyses of liquid RADIOACTIVE EFFLUENT releases during the previous calendar month and the j results of analyses performed on composite samples shall be added to the monthly dose calculation.

A dose tracking system and administrative dose limits shall be established and maintained. Operating parameters shall be adjusted in accordance with methodology described in the ODCM such that the dose values at any time, when projected to the end of the applicable time period, do not exceed the doses g

< specified in Technical Specification 3.17.2.

Bases 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. Yhe ACTION statements provide the required operating flexibility and, at the same time, implement the guides set forth in Section IV.A of Appendix I which 155> assures, by definition, that the releases of radioactive material in liquid effluents will be kept "as low as reasonably achievable." The dose 4 calculations methodology in the 00CM implements the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual 1

155><

Proposed Amendment No. 155 4-72

RANCHO SECO UNIT 1

TECHNICAL SPECIFICATIONS Surveillance Standards TABLE.4.21-1 (Continued)

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM 155> T = [1-exp(-Ats)exp(-Atc)3 A

where x = decay constant (seconds -1)

< te = time from collection to start of counting

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

155>< Prior to sampling, each batch will be isolated, and then thoroughly mixed, to assure representative sampling.

c. Other peaks which are measureable and identifiable, together with the 155> listed nuclides, shall also be identified and reported. Nuclides which are not observed for the analysis shall be reported as "less than" the instrument's LLD, and shall not be reported as being present. The "less than" values shall not be used in the ODCM evaluations. However, if the nuclide is measured and identified at a value less than the Table 4.21-1

< LLD value, it shall be reported and entered into the ODCM evaluations.

155>< d. Miscellaneous Water Evaporator release is via the gaseous pathway.

\

l

. .i q

I Proposed Amendment No. 155 4-71a o

__._.-.__________m

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Dose to Man from Routine Releases of Reactor Effluent for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.13, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

155> The results from composite samples during the period 1981 through 1984 indicates that Cs-137, Cs-134, Co-58 and Co-60 constitute 80 percent of the historical mix of gamma emitting radionuclides in plant liquid effluents.

Another 13 percent consists of I-131. When the thyroid is separated as a limiting organ, 97.8 percent of the total body dose and 97.6 percent of the ifmiting organ dose are due to Cs-134 and Cs-137. Essentially 100 percent of the thyroid dose is due to I-131.

The activity analysis of Cs-134, Cs-137 and I-131 at the Lower Limits of Detection specified in Table 4.21-1 are based on an estimated annual plant ;

radioactive effluent outflow of 20 million gallons per year with an average 1 dilution flowrate of 5,000 gallons per minute. These Lower Limits of Detection provide an adequate basis for determining the presence or absence of dose due to other . radionuclides in plant liquid effluents, when no other indications are revealed during sample analysis.

n' The dose tracking system ensures that the dose limits prescribed in Technical Specification 3.17.2 will not be exceeded at the 95 percent confidence level.

The methodology presented in the ODCM provides for adjustment of operational and analysis parameters to factor in variables such as annual radiological liquid effluent release volume, discharge canal flow rate, and current cumulative dose.

The dose tracking system provides for prompt updating of cumulative dose and contains feedback mechanisms to assure that the target dose values are not exceeded. The tracking system also contains review 'and approval of batch radiological liquid effluent releases at multiple management levels.

There is also reasonable assurance that the operation of the facility will not result in radionuclides concentrations in finished drinking water that are in

< excess of the requirement of 40CFR141.

Proposed Amendment No. 155 4-72a 155x

/

155>

RANCHO SECO UNIT 1

~~

TECHNICAL SPECIFICATIONS' Surveillance Standards

. Table 4.21-2 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM hampling . Minimum Analysis Type of Activity Lower Limit i Liquid Release Frequency Frequency Analysis (c) Of Detection Type (LLD)

(uCi/ml)(a).

Each Batch Each Batch Cs-134 4 x 10-8 A. Batch Waste (gg- Cs-137 5 x 10-8 lease Tanks >

P P I-131 1 x 10-7 H-3 1 x 10-S l

Each Batch Composite (d)- H3 1 x 10-6 P M Na24 3 x 10-5 Cr51 3 x 10-5 Mn54 i x 10-7 Fe59 4 x 10-7 CoS7 4 x 10-6 CoS8 1 x 10-6 Co60 1 x 10-7 Zn65 5 x 10-7 Sr89. 3 x 10-8 Sr90 3 x 10-8

\ Zr95 6 x 10-7 Nb95 1 x 10-8 Mo99 4 x 10-5 t Ag110m 1 x 10-7 1131 '1 x 10-8 1133 5 x 10-6 Cs134 9 x 10-9 Cs136 4 x 10-8 Cs137 1 x 10-8 Ba140 4 x 10-8 La140 6 x 10-6 Co141 5 x 10-7' Ce144 5 x 10-7 Gross Alpha 1 x 10-7 Proposed Amendment No. 155 4-72b i

155>

RANCHO SECO UNIT 1 TECHNICAL

SPECIFICATIONS Surveillance Standards TABLE 4.21 (Continued)

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation '

a. (1) The lower limit of detection (LLD) for a radionuclides presented in this table is the largest concentration, expressed in microcuries per milliliter, which is required to be detected, if present, in order to achieve compliance with the limits Of Specification 3.17.2 (10CFR50, Appendix I).

(2) The LLD of a radioanalysis system is that value which will indicate the presence or absence of radioactivity in a sample when the probability of a false positive and of a false negative determination is stated. The probabilities of false positive and false negative are taken as equal at 0.05. The equation for LLD in microcuries per milliliter is given by the equation:

l LLD = f[2.7 + 3.29(Br)-0.5) 3.7E4(YEVT)-

\

where 2.7 = factor to correct for Poisson distribution at very low background count rates.

W f = correction factor to account for systematic errors = 1.1.

B = background (counts) r=1+

t t

s (t3<t)b .

b tb = background count time (seconds) ts = sample count time (seconds) 3.7E4 = disintegrations /second/ picocurie 1

Y = yield of radiochemical process E = counting efficiency (disintegrations / count)

V = sample volume (liters) or mass (kilograms) 4 Proposed Amendment No.155 4-72c

155>

RANCHO SECO UNIT 1 TECHNICAL SPECIFICAT. IONS Surveillance Standards TABLE 4.21-2 (Continued)

~

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Table Notatior.

T = [1.exp(-Atc)exp(-Atc)]

A where A = decay constant (seconds -1) tc = time from collection to start of counting

b. A batch release is the discharge of iiquid wastes of discrete volume.

Prior to. sampling, each batch will be isolated, and then thoroughly mixed, to assure representative sampling,

c. Other peaks which are measureable and identifiable, together witn the listed nuclides, shall also be identified and reported. Nuclides which are not observed for the analysis shall be reported as "less than" the instrument's LLD, and shall not be reported as being present. The "less than" values shall not be used in the ODCM evaluations. However, if the-nuclide is measured and identified et a value less than the Table 4.21-1 LLD value, it shall be reported and entered into the ODCM evaluations, d A composite sample is one in which the quantity of liquid samples is proportional to\the quantity of liquid waste discharg'ed and in which the method of sampling employed results in a specimen which is representative of the liquids released.

y; ;

i i

9 l

l Proposed Amcndment No. 155 l 4-72d )

< l e

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.21.3 Liquid Holdup Tanks

Surveillance Requirements The quantity of radioactive material contained in each tank listed in Specification 3.17.3 shall be deternined to be within the specified limit by analyzing a representative sample of the tank's contents at least weekly when radioactive materials are being added to the tank.

Bases Restricting the quantity of radioactive material contained in the specified outdoor tanks provides assurance that in the event of an uncontrolled release

' 155> of the tank's contents, the concentration at the nearest potable water supply 4 and the surface water supply in an unrestricted area would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2.

l l

\

4

Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system or the LIQUID EFFLUENT RADWASTE TREATMEi4T SYSTEM.

Proposed Amendment No. 155 4-73

RANCHO SECO UN1T 1 TECHNICAL' SPECIFICATIONS Surveillance Standards 155> 4.21.4

Liquid Effluent Radwaste Treatment Surveillance Requirements Doses due to liquid releases to unrestricted areas shall be projected at least once per 31 days in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM) when LIQUID EFFLUENT RADWASTE TREATMENT SYSTEMS are not being fully utilized. The installed LIQUID EFFLUENT RADWASTE TREATMENT SYSTEN shall be considered OPERABLE by meeting Specifications 3.17.1 and 3.17.2.

Bases The OPERABILITY of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified prov' des assurance that the release of radioactive materials in liquid effluents will be kept "as low as is reasonable achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section 11.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as the dose design objectives set forth in Section II.A of Appendix 1,10 CFR Part 50, for liquid effluents. \ .

Ine installation of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM is not complete. This specification will become effective when the system is declared operable.

Proposed Amendmen,t No. 155 4-73a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standcrds 4.22 GASEOUS EFFLUENTS 4.22.1 Dose Rate Surveillance Requirements 155> The dose rate due to noble gases in gaseous effluents shall be determined to be within the limits in Specification 3.18.1 in accordance with the 4 methodology described in the 0FFSITE DOSE CALCULATION MANUAL (00CM).

The noble gas effluent continuous monitors, as listed in Table 3.16-1, shall use monitor setpoints to limit the dose rate in unrestricted areas to the 155>< limits in Specification 3.18.1.

The release rate of radioactive materials, other than noble gases, in gaseous effluents shall be detennined by obtaining representative samples and performing analyses in accordance with the sampling and analysis program, specified in Table 4.22-1.

155> The dose rate due to Iodine-131, Iodine-133, tritium, and all radioactive material in particulate form with half-lives greater than 8 days released in gaseous effluents, shall be determined to be within the limits in Specification 3.18.1 by using the results of the sampling and analysis program specified in Table 4,22-1, and in accordance with the methodology described in 4 the ODCM. \

J!

Bases This specification is provided to ensure that the dose rate at any time at the 155> Exclusion Area Boundary (Figure 5.1-1) from gaseous effluents 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 concentra'tions of 10 CFR Part 20, Appendix B. Table II, Column 1. These limits provide reasonable assurance 155> 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 B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)(1)). For individuals who may at times be within the Exclusion Area Boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above 155> background to an individual at or beyond the Exclusion Area Boundary to less

< than or equal to 500 mrem / year to the total body or to less than or equal to 3,000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via 155> the grass-cow-milk-infant pathway to less than or equal to 1,500 mrem / year for

< the nearest dairy cow to the plant.

Proposed Amendment No.155 4-74

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.22-1 RADI0 ACTIVE GASEudS WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Analysis Type of Activity Lower Limit ;

Ga:;eous Release Frequency Frequency Analysis of Detection Type (LLD) a (uCi/ml) 155>< A. Waste Gas P P Storage Tank Each Tank Each Tank Principal Gama 1 x 10-4 Grab Emitters (f)

Sample 155> B. Reactor Building P P ;

Purge Vent Each Purge Each Purge (b,e,1) Prin::1 pal Gama 1 x 10-4 l Grab Emitters (f)

< Sampie(b,e,1)

H-3 1 x 10-6 155>< C. Auxiliary M(b,c.e) M(b) Principal Gama 1 x 10-4 Building Stack, Grab Emitters (f)

{ ample H-3 1 x 10-6 155> D. Auxiliary M(b) M(b) Principal Gama 1 x 10-4 Building Grade Grab Emitters (f)

Level Vent Sample H-3 1 x 10-6 155>< E. All Release Continuous W(d) I-131 1 x 10-12 Types as listed Charcoal 155>< in A,B.C,D above Sample I-133 1 x 10-10 Continuous W(d)

Particulate Principal Gama 1 x 10-11 '

Sample Emitters (f)

(I-131, Others) ..

155> Continuous it Gross Alpha (h) 1 x 10-21 Composite Particulate Sample 1

< Sr-89, Sr-90(g) 1 x 10-11 Continuous Noble Gas Noble Gases 1 x 10-6 155>< Monitor Gross Beta and Gama as Xe-133 Proposed Amendment No. 155 4-75

RANCHO SECO UNIT 1 TECHNICAL SPECIFICAT10NS Surveillance Standards Table 4.22-1 (Continued) .

I i

RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Tabb Notation 155> a. ' (1) The lower limit of detection (LLD) for a radionuclides presented in this table is the largest concentration, expressed in microcuries per milliliter, which is required to be detected, if present, in order to achieve compliance with the limits of Specifications 3.18.1, 3.18.2 and 3.18.3.

(2) The LLD of a radioanalysis system is that value which will indicate the presence or absence of radioactivity in a sample when the probability of a false positive and of a f alse negative determination is stated. The probabilities of false positive and false negative are taken as equal at 0.05. The equation for LLD in microcuries per milliliter is given by the equation:

i LLD = f[2.7 + 3.29(Br)-0.53 3.7E4(YEVT) where 2.7 = factor to correct for Poisson distribution at very low background count rates.

\ f = correction factor to account for systematic {

errors = 1.1. f i

4 B = background (counts) i r=1+ t t

s (ts<t)b '

i b

tb = background count time (seconds) ts = sample count time (seconds) 3.7E4 = disintegrations /second/ picocurie Y = yield of r' radiochemical process E = counting efficiency (disintegrations / count) l V = sample volume (liters) or mass (kilograms)

T = [1-exp(- Ate)exp(- Atc)]

A where A = decay constant (seconds -1) tc = time from collection to start of counting 4

Proposed Amendment No. 155 l 4-76 i

. m - __________________._1________________J

RANCHO SECO UN1T 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.22-1 (Continued)

RADI0 ACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation

b. Analysis shall also be performed when gross beta or gamma activity anaylsis of reactor coolant indicates greater than 10 uCf/ml and after each 10 pCi/mi increase in the gross beta or gamma activity analysis.

c. Tritium grab samples shall be taken at least once per seven days from the ventilation exhaust from the auxiliary building stack during refueling 4 and anytime fuel is in the spent fuel pool and the pool temperature exceeds 110*F. Below 110*F there is essentially no evaporation from this i source.

155>< d. Samples shall be changed at least weekly and analyses shall be completed j

within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Sampling and analysis shall also be performed when reactor coolant indicates 10pCi/ml gross beta gamma activity and every l 10pC1/ml increases thereafter. When samples collected for less than 24 I hours are analyzed, the corresponding LLD's maybe increased by a factor of 10.

e. Tritium grab samples shall be taken at least daily during refueling activities. \

j; 155> f. Principle gamma emitters for which the LLD applies are: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-135m and Xe-138 for gaseous samples and Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, (or Tc99m), Cs-134, Cs-137, Ce-141, and Ce-144 for particulate samples. This list does not mean only these nuclides will be detected and reported. Other peaks that are measurable and identifiable shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report, pursuant to Specification d 6.9.2.3. Nuclides which are below the LLD for the analysis shall De reported as "less than" the nuclide's LLD and shall not be reported as being present at the LLD level for that nuclide. However, if the nuclide is measured and identified at a value less than its predetermined LLD value, it shall be reported and entered into the ODCM evaluations.

g. Gross beta analysis performed on a monthly basis for each environmental release particulate sample. . If any one of these samples indicates greater than 1.0 E-11 pCi/cc gross beta activity then a Sr-89, Sr-90 analysis will be perfomed on those samples exceeding this value. j

h. Gross beta perfomed on a monthly basis for each environmental release particulate sample. This fulfills the requirements of performing a !

monthly composite. .

1. After purging seven reactor building volumes, sampling and analysis of f '

Reactor Building Purge Vent exhaust shall be conducted at least once per seven days.

Proposed Amendment No. 155 l 4

4-76a f

- o

155>

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 155>< 4.22.2 Dose-Noble Gases .

Dose Calculations 155> Cumulative air dose contributions for the calendar quarter and calendar year shall be determined in accordance with the methodology described in the OFFSITE DOSE CALCULATIONAL MANUAL (0DCM) at least

< monthly.

Bases This specification is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that 155> conformance with the guides of Appendix I be shown by calculational 4 procedures based on models and data such that the actual exposure of an individyal 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

155>< of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of '

Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," ~ Revision 1, Octoter 1977 and Regulatory Guide 1.111, " Methods I for Estimating Atmospheric Transport and Dispersion of Gaseous l Effluents in Routine Releases from Light-Water-Cooled Reactors," l 155> Revision 1 July 1977. The ODCM equations provided for determining the air doses at the Site Boundary for Gaseous Effluents (Figure 5.1-3) and are based upon the historical average atmospheric

< conditions.

Proposed Amendment No. 155 4-77 i

- .________-______ - _ O

RANCHO SECO UN1T 1 TECHNICAL SPECIFICATIONS Surveillance Standards 155> 4.22.3 Dose-Iodine-131, Iodine-133, Tritium, and Radioactive Materials in Particulate

< Fonn.

Dose Calculations 155> Cumulative dose contributions for the calendar quarter and calendar year p(eriod shall be 00CM) 0FFSITE determined DOSE in accordance CALCULATION with MANUAL at least the methodology described in monthly.

Bases 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 1. The ACTION statements' provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be 155> substantially underestimated. The ODCM calculational methods for calculating

< the doses due to the\ actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109,

" Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1. July 1977. These equations 155>< also provide for estimating doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, and 155>< radioactive material in particulate form are dependent on the existing radionuclides pathways to man at or beyond the Site Boundary for Gaseous Effluents (Figure 5.1-3). 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.

Proposed Amendment No. 155 4-78

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 155>< 4.22.4 Gaseous Radwaste Treatment Surveillance Requirement 155> Doses due to gaseous releases to areas at and beyond the Site Boundary For I Gaseous Effluents (see Figure 5.1-3) shall be projected at least once per 31 !

days in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (0DCM) when Gaseous Radwaste Treatment Systems are not i being fully utilized. {

i The installed VENTILATION EXHAUST TREATMENT SYSTEM and Waste Gas System shall l

< be considered OPERABLE by meeting Specifications 3.18.1, 3.18.2 and 3.18.3. j 1

Bases j 155> The operability of the Waste Gas System and the VENTILATION EXHAUST TREATMENT SYSTEMS ensures that the systems will be available for use whenever gaseous !

effluents require treatment prior to release to the environment. The i'

< requirement that the appropriate portions of systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives i given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits j governing the use of\ appropriate portions of the systems were specified as '

155>< the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents. .

I l

i l

Proposed Amendment No. 155 4-79 ,

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 155>< 4.22.5 Gas Storage Tanks Surveillance Requirements The quantity of radioactive material contained in each waste gas decay tank 155>< shall be determined to be within the limit in Specification 3.18.5 at least I daily when radioactive materials are being added to the tank and the Reactor Coolant System activity exceeds the limits of- Specification 3.1.4.

Bases Restricting the quantity of radioactivity contained in each waste gas decay tank provides assurance that in the event of an uncontrolled release of the 155> tank's contents, the resulting total body exposure to an individual at the

< exclusion area boundary (see Figure 5.1-1) will not exceed 500 mrem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure."

Calculations have shown that the reactor coolant activity must exceed the limits of Specification 3.1.4 before the waste gas decay tank activity 155>< approaches the limits of Specification 3.18.5.

\

4 Proposed Amendment No. 155 4-80 0

l 1

RANCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS l Surveillance Standards 4.25 SOLID RADI0 ACTIVE WASTES Surveillance Requirements 155> 4.25.1 The solid radwaste systems shall be demonstrated OPERABLE at least once per 92 days by:

a. Operating the solid radwaste system at least orce in the ,

previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or

b. Verification of the existence of a valid contract for '

SOLIDIFICATION to be performed by a contractor in ;

accordance with a PROCESS CONTROL PROGRAM.

< 4.25.2 The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at lease one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g.,

filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOL 7.01FICATION parameters can be determined in accordance with che PROCE.% CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may g

then be resumed using che alternative SOLIDIFICATION parameters determined by'the N0 CESS CONTROL PROGRAM.

b. If the initial test specimen from.a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.15, to assure SOLIDIFICATION of subsegarat batches of waste.

155><

Bases 155> The OPERABILITY of the solid radwaste system ens'ures that the system will be available for use whenever solid radwastes require processing and packaging

< prior to being shipped offsite. !

i This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

j Proposed Amendment No. 155 4-81 1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.25 (Continued)'

Bases t 155> The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging

< prior to being shipped offsite.

\

1 Proposed Amendment No. 155 4-82 s., <

b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.26 RADIOLOGICAL ENVIRONMENTAL MONITORING Surveillance Requirements The radiological environmental monitoring samples shall be collected per Table 155> 3.22-1 from the locations shown in the RADIOLOGICAL ENVIRONMENTAL MONITORING

< PROGRAM (REMP) MANUAL and shall be analyzed to the requirements of Tables 3.22-1 and 4.26-1, Bases 155>< The Radiological Environmental 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 implementsSection IV.B.2 of Appendix I to 10CFR50 and supplements the radiological effluent monitoring program by verifying that the measurable concentration of radioactive materials and levels of radiation are not higher than expected on the basis of 155>< the effluent measurements and ODCM modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The specified monitoring program is in effect at the present time. Program changes may be initiated based on operational experience and chang s in regional population or agricultural practices. The 155>< sample locations have been listed in the REMP MANUAL to retain flexibility for making changes as needed.

3 155> The detection capabilities required by Table 4.26-1 are state-of-the-art for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141.

l Proposed Amendment No. 155 4-83 J

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1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.26-1 (Continued)

~ MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD Table Notation

a. (1)

The lower limit of detection (LLD) for a radionuclides presented 155> in this table is the largest concentration, expressed in microcuries per milliliter, which is required to be detected, if present, in order to achieve compliance with the applicable regulation, given stated operating conditions and calculation methology.

(2) The LLD of a radioanalysis system is that value which will indicate the presence or absence of radioactivity in a sample when the probability of a false positive and of a false The probabilities of f alse negative determination is stated. The positive and false negative are taken as equal at 0.05.

equation for LLD in microcuries per milliliter is given by the equation:

LLO = f[2.7 + 3.29(Br)-0.5] _

0.037tYtJIJ _

\ where 2.7 = factor to correct for poisson distribution at very low background count rates.

k f = correction factor to account for systematic errors = 1.1. 3 B = background (' counts) .

r=1+

t s (ts<t}b b

t b;= background count time (seconds) ts:= sample count time { seconds) 0.037 i disintegrations /second/ picocurie Y = yield of radiochemical process E = counting efficiency (disintegrations / count) l V = sample volume (liters) or mass (kilograms) 1 4

Proposed Amendment No. 155 4-85 I

- - - _ - - - _ _ _ - - _ - _ _ - _ _ - _ ____-.____________J

I RANCHO.SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.26-1 (Continued)

MAXIMUM VALUES FOR THE LOWER LIMITS'0F DETECTION (LLD)a, d Table Notation l 155>

.T = [1-exp(-itc)exp(-Atc)]-

A where A = decay constant (seconds -1) ,

tc = time from collection to start of counting (3) Analyses shall be performed in such a manner that the stated LLD's' l will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small semple sizes, the presence of interfering nuclides, or other uncontrollable circumstances may

. render these LLD's unachievable. In such cases, the contributing factors will be toentified and described in the Annual Radiological Environmental Operating Report.

b. LLD for drinking water.

\ 1

c. 'LLD sgown ig for composite analysis. For individual samples, 5x10 pCi/m is the LLD.

g

d. Other peaks which are measurable and identifiable, together with the nuclides in Table 4.26-1, shall be identified and reported.

155>< e. Total for parent 'and daughter.

Proposed Amendment No. 155 4-85a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS-Surveillance Standards 4.27 LAND USE CENSUS Surveillance Requirements The land use census shall be conducted; annually by using methods that will-provide the best results, such as door-to-door survey, aerial survey, or by consulting local agriculture authorities.

155> The land use census or portions .thereof, shall be conducted during the

< appropriate time of the year to provide the best results.

Reports The results of the land use census shall be included in the Annual Radiological Environmental Operating Report.

Bases This specification is provided to ensure that changes in the use of '

155> unrestricted areas are identified and that modifications to the RADIOLOGICAL J

< ENVIRONMENTAL MONITORING PROGRAM MANUAL and ODCM are made if required hy 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 squgre 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 k.. leafy vegetable assumed in Regulatory Guide 1.109 for consumption hy a child.

To determine this minimum garden size, the following assumptions were used:

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

155> In addition, by gathering information on the liquid effluent pathway and the gaseous effluent pathway, the census will assure that proper radiological environmental monitoring and radioactive effluent controls are in place for

< the adequate protection of the health and safety of the general public. i Proposed Amendment No.155 4-86 .

f D

,]

RANCHO SECO UNIT 1 '

TECHNICAL SPECIFICATIONS Surveillance Standards 4.29 FUEL CYCLE DOSE Surveillance Requirements Cumulative dose contributions from liquid and gaseous effluents shall be 155> determined in accordance with Specifications 4.21.2, 4.22.2, and 4.22.3 and in

< accordance with the OFFSITE DOSE CALCULATION MANUAL (0DCM).

155> Cumulative dose contributions from direct radiation (including outside storage tanks, etc.) shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM). This requirement is applicable only under

< conditions set forth in the Action Statement of Specification 3.25.

Reports Special reports shall be submitted as required under Specification 3.25.

Bases This specification is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the numerical 155> guides for design objective doses of Appendix I or exceeds the reporting levels for the Radiological Environmental Monitoring Program. For the Rancho Seco site, it is unlikely that the resultant dose to a MEMBER OF THE PUBLIC 2 will exceed the dose limits of 40 CFR 190 if the plant remains within twice the numerical guides for design objectives of 10 CFR 50, Appendix I and if direct radiation (outside storage tanks, etc.) is kept small. The Special Report will describe a course of action which should result in the limitation of the dose to a MEMBER OF THE PUBLIC for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 q miles must be considered. If the dose to any MEMBER OF THE PUBLIC is ]

evaluated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC! staff action is completed. An individual is not considered a MEMBER OF THE PUBLIC during any period in which l he/she is engaged in carrying out any operation which is part of the uranium )

< fue1, cycle.

Proposed Amendment No. 155 4-89 i

i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS H Design Features

5. DESIGN FEATURES .

155> 5.1 SITE The Rancho Seco reactor is located on the 2,480 acres owned by Sacramento Municipal Utility District, 26 miles north-northeast of Stockton and 25 miles southeast of the City of Sacramento, California. The minimum distance t'o the boundary of the exclusion area, as defined in 10 CFR 100.3, shall be 2,100 feet.

5.1.1 Exclusion Area .

I The EXCLUSION AREA shall be shown in Figure 5.1-1.

5.1.2 Low Population Zone The LOW POPULATION ZONE shall be shown in Figure 5.1-2.

5.1.3 Site Boundary For Gaseous Effluents The SITE BOUNDARY FOR GASE0US EFFLUENTS shall be shown in Figure 5.1-3.

5.1.4 Site Boundary For Liquid Effluents The SITE BOUNDARY FOR LIQUID EFFLUENTS shall be shown in Figure

< 5.1-4.

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I I Proposed Amendment No. 155 5-5 i

- - _ _ - - - .J

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls RESPONSIBILITIES (Continued)

h. Performance of special reviews and investigations and reports 138>< thereon as requested by the AGM, Nuclear Power Production.

138> f. Review of the Plant Security Plan and changes thereto.

j. Review of the Emergency Plan and changes thereto.

k. Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE 155> DOSE CALCULATION MANUAL and the RADIOLOGICAL ENVIRONMENTAL

< MONITORING PROGRAM MANUAL. (See Specifications 6.15 and 6.16.)

1. Review of major changes to the Radioactive Waste Treatment Systems '

(Liquid, Gaseous and Solid), and all information required

< by Specification 6.17.

155> m. Review of any accidental, unplanned, or uncontrolled release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence, and the forwarding of these reports to the Nuclear Plant Manager and to the

< MSRC.

AUTHORITY a 6.5.1.7 The Plant Review Committee shall:

138>< a. Recommend in writing to the AGM, Nuclear Power Production approval or disapproval of items considered under 6.5.1.6(a) through 155>< (m) above.

b. Render determinations in writing with regard to whether or not 155>< each item considered under 6.5.1.6(a) through (e), and (1) above constitutes an unreviewed safety question.

c. Provide immediate written notification to the Ch' airman of the Management Safety Review Committee of disagreement between the 138> PRC and the AGM, Nuclear Power Production; however, the AGM, 4 Nuclear Power Production shall have responsibility for resolution of such disagreements pursuant to 6.5.1.1 above. -

RECORDS 6.5.1.8 The Plant Review Committee shall maintain written minutes of each 138> meeting and copies shall be provided to the AGM, Nuclear Power

< Production and the Chairman of the Management Safety Review Committee.

Proposed Amendment No. 138, Rev. 2 Proposed Amendment No. 155 6-5

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 138>< 6.5.4 (Continued) ,

a. The conformance of facility operation to all provisions contained within the Technical Specifications and applicable license conditions at least once per year.

b. The performance, training and qualifications of the District's entire facility technical staff at least once per year.

c. The result of all actions taken to correct deficiencies occurring in facility equipment, structures, systems or methods of operation that affect nuclear safety at least once per six 138>< months for those changes not previously audited.

d. The performance of all activities required by the Quality Assurance Program to meet the criteria of Appendix "B",10 CFR 50, at least once per two (2) years.

e. The Facility Emergency Plan and implementing procedures at least once per two (2) years.

f. The Facility Security Plan and implementing procedures at least once per two (2) years.

g. Any other area of facility operation considered appropriate by 138>< the MSRC, Deputy General Manager, Nuclear or the General Manager.

h. Complian\ce with fire protection requirements and implementing procedures at least once per two (2) years.

1. An independent fire protection and loss prevention inspection and audit shall be perfomed annually utilizing either qualified offsite licensee personnel or an outside fire protection firm.

j. An inspection and audit of the fire protection and loss prevention program shall be perfomed by an outside qualified fire consultant at intervals no greater than three (3) years.

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

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

155> m. The PROCESS CONTROL PROGRAM and implementing procedures for

< processing and packaging of radioactive wastes from liquid systems at least once per 24 months.

155> n. The performance of activities required by the Quality Assurance

< Program for Effluent Control and Environmental Monitoring.

138> Audit reports of reviews encompassed by Section 6.5.4 shall be forwarded to the General Manager, MSRC Chairman, and to the management positions responsible for the areas reviewed within

< 30 days af ter completion.

Proposed Amendment No.138, Rev. 2 Proposed Amendment No. 155 6-10

RANCHO SECO UNIT 1 {

TECHNICAL SPECIFICATIONS Administrative Controls 6.7 SAFETY LIMIT VIOLATION i 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The provisions of 10 CFR 50.36 (c) (1) (1J and 10 CFR 50.72 shall be complied with.

138> b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Director, Nuclear Operations and Maintenance, the AGM, Nuclear Power Production, and the Chairuan of the MSRC shall be notified

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

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

d. The Safety Limit Violation Report shall be submitted to the 138> Commission, the MSRC, and the AGM, Nuclear Power Production,

< within 14 days of the violation.

^

6.8 PROCEDURES 2 6.8.1 Written procedurss shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.

b. Refueling operations,

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation,

f. Fire Protection Procedures implementation.

155> g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

1. RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL implementation.

j. Quality Assurance Program for the Effluent Control and Environmental Monitoring using the guidance of Regulatory Guide

< 4.15, Revision 1, February 1979.

138> 6.8.2 Each procedure of 6.8.1 above and changes thereto shall be reviewed

< and approved as set forth in Specification 6.5.

Proposed Amendment No. 138, Rev. 2 Proposed Amendment No. 155 6-11 f

\ __ _ _

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.8 PROCEDURES (Continued) 6.8.3 Temporary changes to procedure 6.8.1 above may be made provided:

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 whom holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented, reviewed by the PRC and approved by the Plant Superintendent within seven (7) days of implemen-tation.

6.9 REPORTING REQUIREMENTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

Startup Report 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitt following (1) Receipt of an operating license; (2) amendment of the license involving a planned increase in power level; (3) Installation of fuel that has a different design or has been manufactured by a different fuel supplier; and (4)

! modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program.and comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

6.9.1.2 Startup reports shall be submitted within (1) Ninety (90) days following completion of the startup test program; (2) Ninety (90) days following resumption or commencement of commercial power operation; or (3) Nine (9) months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three (3) months until all three events have been completed.

Proposed Amendment No. 155

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 155>< 6.9.2 Radiological Reports 6.9.2.1 Annual Radiological Reports

! Annual reports covering the activities of the unit, as described 155> below, for the previous calendar year shall be submitted as follows:

6.9.2.1.1 Annual Occupational Radiation Exposure Report The Annual Occupational Radiation Exposure Report shall be submitted to the Commission within the first calendar quarter of each calendar year in compliance with 10CFR20.407.

6.9.2.1.2 Annual Exposure Report l

The Annual Exposure Report shall be submitted to the Commission within the first calendar quarter of each calendar year in

< accordance with the guidance contained in Regulatory Guide 1.16.

6.9.2.2 Annual Radiological Environmental Operating Report 155>< 6.9.2.2.1 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar 155>< year shall be submitted prior to May 1 of each year.

155>< 6.9.2.2.2 The Annual Radiological Environmental Operating Reports shall g 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 surveillance reports, and an -

assessment of the observed impacts of the plant operation on the environment. The reports shall aise include the results of the 155>< Land Use Census required by Specification 3.23. If hannful 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.

155><

1 Proposed Amendment No. 155 6-12a 1

)

i

)

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2.2.2 (Continued) 155>< The Annual Radiological Environmental Operating Reports shall 138> include summarized and tabulated results of all radiological

< environmental samples taken during the report period. In the

. event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following: a sumary 155> description of the Radiological Environmental Monitoring

< Program; including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used; a map of all sampling locations keyed to a table giving distances and directions from one reactor; the result of land use censuses, and the results of licensee participation in the Interlab Comparison Program. The annual report shall also 155> include information related to Specification 4.29, Uranium Fuel

< Cycle Dose.

6.9.2.3 Semiannual Radioactive Effluent Reiease Report 155>< Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous six months of operation shall be submitted within 60 days after January 1 and July 1 of each year.

155><

155>< 6.9.2.3.1 The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Re'1 eases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized on a quarterly basis, following the format of Appendix B thereof.

155> The Semiannual Radioactive Effluent Release Report shall include a summary of hourly meteorological data collected over the

< report period. .

Proposed Amencent No.138, Rev. 2 Proposed Amendment No. 155 6-12b 3

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2.3.1 Continued) 15 5'> < The Semiannual Radioactive Effluent Release Reports shall include an assessment of the radiation dosos from radioactive i

gaseous and liquid effluents to individuals due to their i l

155>< activities inside the sits boundary (Figures 5.1-3 and 5.1-4) )

during the report period. Ali assumptions used in making these assessments (e.g., specific activity, exposure time, and ;

location) shall be included in these reports. j i

155>< The Semiannual Radioactive Effluent Release Reports shall j include the following information for all unplanned releases to {

unrestricted areas of radioactive materials in gaseous and I liquid effluents:

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 radioactive 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. The assessment of radiation doses shall be performed in accordance with the 155> 0FFSITE DOSE CALCULATION MANUAL (ODCM).

The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP), RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL and 0FFSITE DOSE CALCULATION MANUAL (0DCM) pursuant to Specifications 6.15 and 6.16 as well as any major changes to Liquid, Gaseous or Solid Radwater Treatment Systems pursuant to Specification 6.17.

The Semiannual Radioactive Effluent Release Report shall include tables for comparison with Specifications 3.17.2, 3.18.2, and 3.18.3. The July-December report shall include a summary table for comparison with the annual values:in Specifications 3.17.2, 3.18.2, and 3.18.3.

The Semiannual Radioactive Effluent Release Report shall also include events described in Specifications 3.17.1, 3.17.3, 3.18.1 and 3.20.

Proposed Amendment No. 155 6-12c

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2.3.1 Continued)

~

155> The Semiannual Radioactive Effluent Release Report shall include the following information for each type of solid waste shipped offsite during the report period:

a. Container volume,

b. Total curie quantity (determined hy measurement or estimate),

c. Principal radionuclides (determined by measurement or estimate),

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

e. Type of container (e.g., LSA, Type A, Type B, High Integrity),and

f. Solidification agent (e.g., cement).

MONTHLY REPORT 6.9.3 Routine reports of operating statistics, including narrative summary of operating and shutdown experience, of lifts of the Primary System Safety Valves or EMOVs, of major safety related maintenance, and t tabulations of facility changes, tests or experiments required pursuant to 10 CFR E0.59(b), shall be submitted on a monthly basis to the Office of Director of Inspection and Enforcement, U. S. huclear Regulatory Commission, Washington, D. C. 20555, with a copy to the Regional Office, postnarked no later than the 15th day of each month following the calendar month covered by the report.

LICENSEE EVENT REPORT 6.9.4 The LICENSEE EVENT REPORTS of Specification 6.9.4.1 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC as Licensee Event Reports. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a License Event Report shall be completed and reference shall'be made to the original report date, pursuant to the requirements of 10 CFR 50.73.

(

Proposed Amendment No. 155 6-12d

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls gCENSEEEVENTREPORT 6.9.4.1 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty (30) ,

days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form, pursuant to 10 CFR 50.73 and the guidance of NUREG-1022.

a. (i) The completion of any nuclear plant shutdown required by the plant's Technical Specification; or (ii) any operation or condition prohibited by the plant's Technical Specifications; or (iii) Any deviation from the plant's Technical Specifications authorized pursuant to 10 CFR 50.54(x).

b. Any event or condition that resulted in the condition of the ,

r.uclear power plant, including its principal safety barriers, '

being seriously degraded, or that resulted in the nuclear power plant being:

(1) In an unanalyzed condition that significantly compromised plant safety; (ii) In a condition that was outside the design basis of the pihnt; or ,

i

< (iii) In a condition not covered by the plant's operating '

and emergency procedures.

c. any natural phenomenon or other external condition that posed an actual threat to the saf.ety of the nuclear power plant or i significantly hampered site personnel in the performance of l duties necessary for the safe operation of the nuclear power plant.

d. Any event or condition that resul.ted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (RPS). . However, actuation of an ESF, including the RPS, that resulted from and was part of the preplanned sequence during testing or reactor operation need not be reported. :

e. any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to:

1. Shut down the reactor and maintain it in a safe shutdown condition; Proposed Amendment No. 155 155>< 6-12e

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls LICENSEE EVENT REPORT

2. Remove residual heat;

3. Control the release of radioactive material; or

4. Mitigate the consequences of an accident.

f. Events covered in paragraph 6.9.4.1.e of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/cr procedural inadequacies. However, individual component failures need not be reported pursuant to this paragraph if redundant equipment in the same system was operable and available to perform the required safety function.

g. Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

1. Shut down the reactor and maintain it in a safe shutdown condition;

2. Remove residual heat; i

\

3. Control the release of radioactive material; or I

" 4. Mitigate the consequences o'f an accident.

h. 1. Any airborne radioactivity release that exceeded 2 times the applicable concentrations of the limits specified in Appendix B, Table II of 10 CFR 20 in unrestricted areas, -

when averaged over a time period of one hour.

2. Any liquid effluent release that exceeded 2 times the 1 limiting combined Maximum Permissible Concentration (MPC)

(see Note 1 of Appendix B to 10.CFR 20) at the point of entry into the receiving water (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases, when averaged over a time period of one hour. l

1. Any event that posed an actual threat to the safety of the i nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nucle 6r power plant including fires, toxic gas releases, or l radioactive releases. l l

j. Failure of the pressurizer EMOVs or Primary System Safety Valves.

Proposed Amendment No. 155 155x 6-12f {

1

)

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls Special Reports 6.9.5 Special reports shall be submitted to the Regional Administrator, Region Y 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 one-time only, " Narrative Summary of Operating Experience" A.

will be submitted to cover the transition period (calendar year 1977).

B. A Reactor Building Structural integrity report shall be submitted within ninety (90) days of completion of each of the following tests covered by Technical Specification 4.4.2 (the integrated leak rate test is covered in Technical Specification 4.4.1.1).

1. Annual Inspection

2. Tendon Stress Surveillance

3. End Anchorage Concrete Surveillance l

4. Liner Plate Surveillance C. Inserv(ce Inspection Program D. Inoperable Accident Monitoring Instrumentation 30 days (3.5.5)

E. Status of Inoperable Fire Protection Equipment F. Inoperable Emergency Control Room /TSC Ventilation Room Filter System i G. Radioactive Liquid Effluent Dose 30 days (3.17.2)

H. Noble Gas Limits 30 days (3.18.2)

I. Radiofodine and Particulate 30 days (3.18.3) 155> J. Gaseous and Liquid Radwaste Treatment 30 days (3.18.4 and 3.17.4)

< K. Radiological Environmental Monitoring Phogram 30 days (3.22)

L. Monitoring Point Substitutions 30 days (3.22J 165>< M. Solid Radioactive Wastes 30 days (3.21)

N. Fuel Cycle Dose 30 days (3.25) 155>< 0. Land Use Census 30 days (3.23)

P. Steam Generator Tube Inspection 30 days (4.17.5)

Proposed Amendment'No 155 155>< 6-129

_2

RANCHO SECO UN2T 1 TECHNICAL SPECIFICATIONS Administrative Controls

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

c. Records of facility radiation and contamination surveys.

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

e. Records of gaseous and liquid radioactive material released to !

the environs, i

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

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

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

1. Records of Quality Assurance activities required by the QA Manual,

j. Records of reviews performed for changes made to procedures or equipmen,t or reviews of tests and experiments pursuant to 10 CFR 50.59.

I k. Records of meetings of the PRC and the MSRC.

1. Records for Environmental Qualification which are covered under the provisions of paragraph 6.14.

155x m. Records for the Radiological Environmental Monitoring Program.

n. Records of the maintenance of all hydraulic snubbers listed in Table 3.12-1.

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.

6.12 - Deleted Proposed Amendment No. 155 6-14

P RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.15 PROCESS CONTROL PROGRAM (PCP) 6.15.1 Function 155> The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured.

6.15.2 Changes A. The PCP shall be approved hy the Commission prior to implementation.

B. Licensee initiated changes to the PCP shall:

1. Be submitted to the Commission by inclusion in the

< Semiannual Radioactive Effluent Release Report for the period in which the change (s) was/were made and shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; b.\ A determination that the change did not reduce the overall conformance of the solidified waste product to 1 existing criteria for solid wastes; and

c. Documentation of the fact that the change has been reviewed and found acceptable by the Plant Review Committee.

2. Become effective upon review and acceptance by the PRC, unless otherwise acted upon by the Commission through written notification to the Licensee.

Proposed Amendment No. 155 6-17

' RANCHO SECO UNIT 1 d TECHNICAL SPECIFICATIONS i i

Administrative Controls )

1 155>< 6.16 0FFSITE DOSE CALCULATION AND RADIOLOGICAL ENVIRONMENTAL MONITORING {

PROGRAM MANUALS l

6.16.1 Function 1 1

155> 6.16.1.1 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall describe the methodology and parameters to be used in'the calculation of offsite

< doses due to the release of radioactive material in gaseous and .

liquid effluents and in the calculation of gaseous and liquid I effluent monitoring instrumentation alarm / trip setpoints consistent )

with the applicable LC0's contained in these Technical Specifications. Methodologies and calculational procedures ;

acceptable to the Commission are contained in various Regulatory Guides as noted in the bases of applicable LC0's.

6.16.1.2 The RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) MANUAL shall be a manual containing the description of the Rancho Seco Radiological Environmental Mcnitoring Program. The REMP manual shall !

contain a description of the environmental samples to be collected, i the sample locations, sampling frequencies, and sample analysis criteria.

6.16.2 Any changes to the ODCM or REMP MANUAL shall be made as follows:

A. Licensee-initiated changes:

2 1. Shall be submitted to the Commission by inclusion in the 155> Semiannual Radioactive Effluent Relecse Report for the

< period in which the change (s) was/were made and shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages 155>< of the ODCM and the REMP MANUAL 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 PRC and MSRC.

2. Shall become effective upon a date specified and agreed to by both the PRC and MSRC following their review and acceptance of the change.

Proposed Amendment No.155 6-18

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.17 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (LIQUID, GASE0US, AND SOLID) 6.17.1 Function The radioactive waste treatment system (liquid, gaseous, and solid) 155>< are those systems described in the facility Updated 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 LCOs set forth in these Specifications.

6.17.2 Major changes to the radioactive waste systems (liquid, gaseous, and 138> solid) shall be made by the following method. For the purpose of this

< specification, " major changes' is defined in Specification 6.17.3.

138> Licensee-initiated changes:

1. The Commission shall be informed of all changes by the inclusion of a suitable discussion of each change in the Semiannual Radioactive 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 determination that the change could be made in accordance with 10 CFR 50'.59; f b. Sufficient information to support the reason for the change without benefit of additional or supplemental information;

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

d. An evaluation of the change with regard to the predicted releases of radioactive materials in liquid and gaseous i effluents and/or quantity of solid waste if different from those previously predicted in the license application and amendments thereto; 155> e. An evaluation of the change with regard to the expected maximum exposures to a MEMBER OF THE PUBLIC in the

< UNRESTRICTED AREA and to the general population if different from those previously estimated in the license application and amendments thereto; I

Proposed Amendment No. 138, Rev. 2 Proposed Amendment No. 155 l 6-19

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE'0F CONTENTS APPENDIX 8 PAGE !

1 1.0 DEFINITIONS 2.0 ENVIRONMENTAL PROTECTION CONDITIONS 4 2.1 Deleted 2.2 Deleted 2.3 Deleted 2.4 Deleted 2.5 Deleted 155H 2.6 Deleted 20 3.0 NON-RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE PROGRAMS 25 3.1 Erosion 25 3.2 Drift Contaminants 25 3.3 Deleted 3.4 Noise 26 3.5 Fogging 27 3.6 Deleted 4.0 RADIOLOGICALENkIRONMENTALMONITORING 31

! 4.1 Deleted 4.2 Deleted 4.3 Deleted 4.4 Deleted 4.5 Deleted ,

4.6 Deleted 4.7 Deleted d i

4.8 Deleted 4.9 Deleted 4.10 Deleted 5.0 ADMINISTRATIVE CONTROLS 42 ,

5.1 Responsibility 42 5.2 Organization 42 ;

5.3 Review and Audit 42 I 5.4 Action to be taken in Event of Violation of an Environmental Protection Limit 43 ,

5.5 Procedures 44 {

5.6 Plant Reporting Requirements 44 5.7 Records Retention 46 5.8 Deleted Proposed Amendment 155 i

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155>

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4 Proposed Amendment No.155

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155>

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4 Proposed Amendment No. 155 .

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