Text

Rev 1 to Proposed Amend 155,revising Radiological Effluent Tech Specs to Incorporate Clarifications & Correct Editorial Errors
	ML20235R405
Person / Time
Site:	Rancho Seco
Issue date:	10/03/1987
From:	Firlit J; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Miraglia F; Office of Nuclear Reactor Regulation
Shared Package
ML20235R409	List: ML20235R405; ML20235R426;
References
	AGM-NPP-87-293, TAC-56033, TAC-64735, TAC-64736, NUDOCS 8710080038
	Download: ML20235R405 (351)
	v • d • e

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$)SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT D P. O. Box 15830, Sacramento CA 95852-1830,(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA l

October.3, 1987 AGM/NPP'87-293 U. S. Nuclear Regulatory Commission Attn: Frank J. Miraglia, Jr.

Associate Director for Projects Philips Bldg.

7920 Norfolk Avenue Bethesda, MD 20014 DOCKET 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION UNIT #1 PROPOSED AMENDMENT NO. 155, REVISION 1

Dear Mr. Miraglia:

By letter to the Commission dated June 30, 1987, the District submitted Proposed Amendment No.155 (PA-155) Revision 0 to the Radiological Effluent Technical Specifications. The District hereby submits Revision 1 of Proposed Amendment No.155 which replaces Revision 0 in its entirety.

Revision 1 provides additional clarifications and corrects editorial errors.

The per batch liquid release analyses are combined into one table (Table 4.21-1) at Lower Limits of Detection (LLDs) which are sufficient for

, compliance with 10 CFR 20 and which provide reasonable assurance of compliance l with the ALARA dose guidelines of 10 CFR 50, Appendix I. The definitions of Lower Limit of Detection have been modified. Attachment A provides an index for this submittal.

Pursuant to 10 CFR 50.91(b)(1), the Radiological Health Branch of the California State Department of Health Services has been informed of this proposed amendment by mailed copy.

Because this is a revision to Proposed Amendment No.155, no additional license fee is required.

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PDR RANCHO sECO NUCLEAR GENERATING STATION O 14440 Twin Cities Road, Herald, CA 95638-9799;(209) 333-2935

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! AGM/NPP 87-293 Frank J. Miraglia, Jr. -

2-If you have any questions concerning this submittal, please contact Mr. Jerry Delezenski at (916) 452-3211, extension 3909.

Sincerely,

/_g2 oseph F. Firlit Assistant General Manager, Nuclear Power Production l Attachments cc w/atch:

G. Kalman, NRC, Bethesda (2)

A. D'Angelo, NRC, Rancho Seco J. B. Martin, NRC, Walnut Creek (2)

Sworn to and subscribed before me this 341 day of October,1987. <

hbu]/b/br Notary Publicgf i

< OFFICIAL SEAL ANNE SHELBY d

NOTARYPUBUC CAUFoRNIA i

SACRAMENTO COUNTY u, co.. e# u, u. mi l, i

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ATTACHMENT A l INDEX TO PA-155 SUBMITTAL 1

ATTACHMENT 1 - Safety Analysis and No Significant Hazards Determination ATTACHMENT 2 - Administrative and Editorial Changes to Rancho Seco Technical Specifications per PA-155.

ATTACHMENT 3 - Bases for Lower Limit of Detection Values for Rancho Seco Liquid Effluents.

ATTACHMENT 4 - PA-155 with List of Effective Pages (changes denoted in right-hand margin).

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I ATTACHMENT 1 l SAFETY ANALYSIS OF PROPOSED AMENDMENT NO. 155 I

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

Proposed Amendment 155 consists of changes and additions to the Rancho Seco Technical Specifications regarding radiological effluents to ensure compliance with the requirements of 10 CFR 50, Appendix I, as it pertains to effluent monitoring, nearby land use identification, offsite dose calculations, environmental monitoring and limitations on calculated offsite doses.

REASON FOR CHANGE:

The amendment proposed in this safety analysis are a result of eteclusions documented in a July 22,1986, " Staff Evaluation of Rancho Seco Reif ol'gical Effluent Technical Specifications and Offsite Dose Calculation Manual Regarding Offsite Doses From Liquid Effluent." This document noted that radionuclides concentrations near the Rancho Seco site resulting from the release of liquid effluents were found to be at levels which could cause the maximum - exposed individual to receive doses well in excess of the dose limits set forth in the 10 CFR 50, Appendix I and 40 CFR 190. Specifically, as a result of the NRC review, an inconsistency was found between the Lower Limit of Detection (LLD) listed in Table 4.21-1 of the Rancho Seco Technical Specification Surveillance Standards (Section 4.21.1) and Technical Specification Sections (3.17.2 and 4.21.2) relating to 10 CFR 50 Appendix I design objectives. Additi,onal NRC citings for nonconformance to 10 CFR 50 Appendix I were documented in an NRC Region V Inspection Report 86-15 sent to the District on June 6, 1986. District response to the citings on July 3, 1986 include commitments to submit Technical Specification revisions for conformance with the requirements of 10 CFR 50 Appendix I. The commitments are addressed in the Proposed Technical Specification Amendment No. 155.

EVALUATION AND BASIS FOR SAFETY FINDINGS The following is a discussion of each Technical Specification change. The changes include administrative changes, changes in accordance with the design basis for Rancho Seco and changes that represent additions or upgrade to existing surveillance requirements and limitations. The Tech Spec changes listed in this analysis will be formatted by individual tech spec changes, accompanied by the existing specification, and a discussion of the change. A safety analysis and a no significant hazards consideration will also be included with each change. In addition, attached are copies of the Proposed Amendment No. 155 of the Technical Specifications.

FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 q PROPOSED AMENDMENT NO. 155 PAGE 2 !

1. Existing Specification:

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1.13 PROCESS CONTROL PROGRAM f A PROCESS CONTROL PROGRAM (PCP) shall be the manual' detailing the program of sampling, analysis, and evaluation 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 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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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:

1.15 0FFSITE DOSE CALCULATION MANUAL (0DCM)

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: ,

d 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 delinngtad and titled the Radiological Environmental Monitoring Program '

(REMI ,iiaual. 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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I FACILITY CHANGE SAFETY ANALYSIS . LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 3a 2a. Existing Specification:

.1.17 SITE BOUNDARY 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.

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 SAFETY 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, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844,

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

New Specification:

1.18 DOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration vf I-131 (microcurie / gram) which alone would produce the same 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, "Likits for Intakes of Radionuclides by Workers," 1979.

Discussion:

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 information available documented in ICRP Publication 30, " Limits for Intake of Radionuclides by Workers," 1979.

FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 j PROPOSED AMENDMENT No. 155 PAGE 5

4. Existing Specification:

N/A New Specification:

1 1.21 MAXIMUM HYPOTHETICAL INDIVIDUAL f 1

The MAXIMUM HYPOTHETICAL INDIVIDUAL is characterized as l

" maximum" with regard to food consumption, occupancy, or j other usage or exposure pathway parameters in the vicinity i 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 l the potential pathways at the " maximum" value. j The MAXIMUM HYPOTHETICAL INDIVIDUAL is a hypothetical receptor of radiological erposure (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 consist.ent 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 PROGRAM (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 l liquid effluents by collecting liquid effluent and providing processing for the , purpose of reducing the total radioactivity prior to release to'the environment.

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i l FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT N0. 155 PAGE 6

4. New Specification: (Cont.)

1.24 VENTILATION EXHAUST TREATMENT SYSTEM The VENTILATION EXHAUST TREATMENT SYSTEMS are systems designed and installed to reduced gaseous radiofodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulate from the gaseous exhaust stream ,

prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEMS components.

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 manner that replacement air or gas is required to purify the confinement.

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 RADI0 ACTIVE EFFLUENT I 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 Tables 4.21-1, 4.21-2 and 4.22-1. [ !

I Discussion:

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 definitions of Maximum Hypothetical Individual (MHI) is necessary to differentiate methodologies for measuring compliance with 10CFR50 Appendix

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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5. Existing Specification:

3.15 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation a channels shown in Table 3.15-1 shall be OPERABLE with their i alarm / trip setpoints set to ensure that the limits of i 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 value 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 acceptably conservative.

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

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 areas 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 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 Regenerant Hold-Up Tanks.

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

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

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5. Existing Specification: (Cont.)

1 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 setpoints 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 (ODOM).

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

4 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, ,

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 f

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5. New Specification: (Cont.)

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.

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 Tcnk (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 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.

6..Existinn Specification: Page '11 Table 3.15-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Number of Channels Instrument Operable- Action

1. Gross Radioactivity Monitors Providing Auto-matic Tennination of Release

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

. 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 l

- release rate calculations and i df scharge valving.

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

2. FTow Rate Measurement Devices

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a. Regenerant Hold-Up Tank 1 With the fTow rate measurement Discharge Line Monitor device inoperable, effluent releases via this pathway may 4 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 i during actual releases. Pump '

perfonnance curves generated in situ may be used to estimate flow.

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6.'_Existinn Specification:'(Cont.) Page 12 ,

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Table 3.15-1 (Continued)

RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION '

Minimum Number of Channels Instrument Operabid ' Acti on -

2. Flow Rate Measurement Devices (Continued)

b. Waste Water Flow 1 With the flow rate measurement device inoperable, effluent releases via this pathway may continu'e 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:

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-Table 3.15 RADI0 ACTIVE LIQUID EFFLUENT WONITORING INSTRlNENTATION 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 from the retention basin: d

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

' 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 Table 3.15-1 (Continued)

RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum ,

Number of Channels Instrument Operable Action

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 4 and Totalizer device inoperable, effluent releases 2hi via this pathway may. continue provided the flow rate is estimated at least once per's hours during retention basin releases.

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FACILITY CHANGE SAFETY ANALYSIS LOG N0. 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 RHUTs A and B total, flow monitor is

'added to assess total curies of radioactive material' released offsite .for offsite dose calculation. The addition of the RHUT Discharge Line_ Monitor

specification is provided for measurement of the RHUT volume released to 3 the retention basin and the determination of total'offsite dose. ]

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PACILITY 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 j channels shown in Table 3.16-1 shall be OPERABLE with their I 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 3.16-1.

Action

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 effluen't 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.

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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 requirements of General Design Criteria 60, 63, and 64 of Appendix 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.

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' FACILITY CHANGE SAFETY ANALYSIS. LOG NO. 921 ,

PROPOSED AMENDMENT NO.155 PAGE 17 l

7. Existing Specification: (Cont.)

The air ejector exhaust and gland seal exhaust also have individual ,

noble gas monitors. These systems exhaust into the Auxiliary .

l 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 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The. radioactive gaseous effluent monitoring instrumentation l channels shown in Table 3.16-1 shall be.0PERABLE with their alarm / trip setpoints set to ensure that the limits of l Specification 3.18.1 (a) are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the methodology contained in the ODCM.

Continuous samples of the gaseous effluent for radioiodines and radioactive particulate material shall be taken as indicated in Table 3.16-1.

Applicability During release via the pathways identified in Table 3.16-1. ,

1 Action j

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 3.18.1 are met, immediately suspend the release of radioactive gaseous effluent i 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, explain in the next Semiannual Radioactive Effluent Report pursuant to Specification 6.9.2.3 why the inoperability was not corrected in a timely manner.

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i FACILITYCHANGESAFETYANkLYSIS LOG NO. 921 PROPOSED AMENDMENT NC. 155 PAGE 18 i , t New Specification: (Cc6t.T 7.

Bases The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in ' gaseous effluents during actnti or potential releases of gaseous ef.fluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the methodology contained in the CDCM to ensure that the alarm / trip will' occur prior to exceeding the limitst 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 terminati m 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 individuy1 voble gas 1 monitor. This system exhausts into the Auxiliary thilding ventilation system. Therefore, the Auxiliary Building Stack is the.

effluent release point and will alarm upon release of environment 7iy significant rladioactive 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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D Table 3.16-1 U

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION

(, .,- c Minimum Number

. i of Chan~nels '

Instrument i . Operable Action _

1. Reactorduilding P'Irge '

Vent. 's ,.

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

Monitor, providing tHp setpoint less conservative alarm and automatic ,

, than required by Specification termination of 2.16, immediately suspend the

.releast e release or declare the channel inoperable. >

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With the monitor inoperable.

effluent releases via tnis 1 4athway 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 24- hours.

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 .

I

c. Particulate Sampler 1 With the collection device inoperable, effluent releases via this pathway may continue i

! provided continuous samples are taken and these samples are analyzed in accordance.with 3 e

Table 4.22-1 within 24. hours.

N

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8. Existing Specificat-lon:_ (Cont.) Page 20

, // s d

Table 3;16-1 (Continued) -

RADI0 ACTIVE GASES EFFl.UENT MONITORD4G INSTRUMENTATION v Minimum [

'% Number ,

of Channels Instrument 4

Operable Action

s s h

{ , ,

l ,

, 1.

Reac$or' Vent (continued Building)Purp

y

d. System Effluent Flow 1 With the flow rate device f inoperable, effluent releases d

Rate Device' may, ce0C(nue providea the flow

, # rate used is the maximum design flow rate.

(

e. 3 Sampler Flow Rate 1. With.ttn flow ra'te device Measurement Device (noptrable, effluent releases-via. tielf.' pathway may continue provfejad the flow rate is estinutad and recorded at least g onctoner 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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l l 8, Existing Specification: (Cont.) Page 21 Table 3.16-1.(continued) ]

i RADIOACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Minimm j Number of Channels Instrument Operable Action l

2. Auxiliary Building Stack

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

Monitor providing trip setpoint less conservative alana and automatic than required by Specification termination of 3.16, imediately suspend the release release or declare the channel inoperable.

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

\ samples are analyzed in accordance with Table 4.22-l g . ..

withftr 7.4. hours.

b. Iodine Sampler I 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 j 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 l taken and these samples are analyzad in accordance with TabTe 4.22-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l

. i l

O O

- ~

}

, 8. Existina Specification:1(Cont. Page 22 l

Table 3.16-1 (Continued)-

)

L RADI0 ACTIVE GASES EFFLUENT MONITORING. INSTRUMENTATION Minimum Number of Channels ;

Instrument Operable _

Action i

2. Auxiliary Building

., Stack (continued)

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

j

e. Sampler Flow Rate 1 With the flow rate device ]-q Measuring Device inoperable, effluent releases via this pathway may continue l 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. Wasta Gas Holdup- 1 With the monitor channel alann/ .

' System ( Auxilf ary trip setpoint less conservative .

Building Stack than required by Specification ?

i Monitor) j 3.16, immediately suspend the release or declare the channel

- inoperab1e.

With the monitor inoperable, the contents of the tank (s) may be _l released to the environment i provided that prior to - ,

initiating the release: l

a. At least two independent 3

' samples of the tank's contents are analyzed, and

b. At least two technically qualiffed members of the ,

Facility Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this

" pathway.

~

?

A

= 4 -

t .

4

8. Existing Specification: .(Cont.) Page 23 Table 3.16-1 (continued)

RADIOACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION l Minimum Number of Channels -

Instrument Operable ,

Action

3. Radwaste Service Area. Vent * .

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

Monitor trip setpoint less conservative than required by Specification 3.16, imediately 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 i within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c

b. Todine Sampier 1 .~ With the co1Tection device

. inoperable, effTuent reTeases y via this pathway may continue provided continuous samples are ~

taken and these samples are .

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

Thei Radwaste Service AM Vent Monttoring System is not yet functional . l This speciffcatfam for this systen will. become effective when it is declared OPERABLE. ,

l

.= .

e

8.LExistina Specification: (Con t .") Pa;;e 24

~ .:

p. .. .

. Table 3.16 (continued)

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION l

Minimus Number of Channeis !

Instrument Operable Action

3. Radwaste Service ,

Aree Vent * (continued) .

With the collection device

c. Particulate Sampler 1 inoperable, effluent releases via this pathway may contir ue 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 maximus design fTow rate. ,

e. . Sampler FTow Rate 1 With the flow rate device Measurement Device inoperable, effluent releases via this pathway may continue y

'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.

4 e

e e e h _ _ _ _ _ _ __ - _ _ . _ . _ _ _ _ . _ . . _ _ -

~

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v i i

l

'8. New Specification: '_Page 25'

.. a . - . . _.

p' Table 3.16-1 RADIOACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION Hinimum Number ,

of Channels

. Instrument Operable Action

1. Reactor Building Purge Vent

a. Noble Gas Activity 1 With the monitor channel alarm / trip Monitor providing setpoint less conservative alarm and automatic than required by Specification termination of 3.18.1, inanediately suspenu the release.

release or declare the channel inoperabl e.

h 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 Tabic 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 coll'ection device

inoperable, effluent releases.via this pathway may continue provided i continuous samples are taken within one hour af ter the monitor is /A; declared inoperable and these N 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 within one hour af ter the monitor is declared inoperable .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 . **

See Table 3.5.5-1 for action required when this monitor is considered an accident monitor.

- Interruption of continuous saupling is allowed for periods not to exceeo '

one hour.

- - - . , - - - - - , - - , - - - - - - - - - - - - - - - - - a

8. New Specification: (Cont.) Page 26 Table' 3.16-1 (Continued)

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

1. Reactor Building Purge Vent (continued)

d. System Ef fluent 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 or the measured flowrate, whichever is greater.

e. Sampler Flow Rate 1 With the flow rate device 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 .

n i

'8.New Specification: (Cont.) Page 27 Table 3.16-1 (Continued)'

L RADI0 ACTIVE 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 provicing effluent releases via this pathway.

alarm

may continue provided grab samples d, 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 within one hour after the monitor is.

declared inoperable and these samples are analyzed in accordance h 1 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 within one hour af ter the monitor is declared inoperable 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

See Table 3.5.5-1 for action required when this monitor is considered un accident monitor.

Interruption of continuous sampling is allowed for periods not to exceed 1 one hour. 1 i

i

d

8. New Specification: Page 28

^

Table 3.16-1 (Continued) '

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 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 or the measured flowrate, whichever is greater, d

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 alarm /

System ( Auxiliary trip setpoint less conservative than required by Specification Building)

Monitor Stack 3.18.1, immediately suspend the 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 samples of the tank's l contents are analyzed, and

b. At least two technically l qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of I radioactive ef fluents via thi s j pathway. l l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

I

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

RADIOACTIVE 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 /fs 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 within one hour af ter the monitor is declared inoperable 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

See Table 3.5.5-1 for action required when this monitor is considered an accident monitor.

- Interruption of continuous sampling is allowed for periods not to exceed one hour.

t

p;

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

_j L.

8. New Specification: (Cont.) Page 30 L

Table 3.16-1 (continued)

RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION - )

Minimum Number

' of Channel s Instrument Operabie Action

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 within one hour af ter the monitor is declared inoperable and these samples are analyzed in accordance with Table 4.22-1 k

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 or the measured n

- flowrate, whichever is greater. lli

e. Sampler Flow Rate 1 With the flow rate device 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 .

- Interruption of continuous sampling is allowed for periods not to exceed one hour.

l' I

1-l FACILITY' CHANGE SAFETY ANALYSIS LOG NO. 921-PROPOSED AMENDMENT NO.155 PAGE 31

8. New Specification 1 (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 i specified in Technical Specification ~ 3.18.1. Additional changes are based l on compliance with Standard RETS, except that the interruption of-continuous iodine and particulate sampling is made allowable for periods not to exceed one hour.

l A note has been added to reference Table 3.5.5-1 for action required when the noble gas activity monitor is considered an accident monitor.

l f'

I i

1

r-_ _ --_

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

i

9. Existing Specification:

3.17 LIQUID EFFLUENTS 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 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 ih 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 Liquid Effluents (see rigure 5.1-4) shall be limited to the concentrations specified in 10CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved

FACILITY CHANGE SAFETY ANALYSIS LOG N0. 921 PROPOSED AMENDMENT NO 155 PAGE 33 i

9. New Specification (Cont.)

or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2X10-4 uC1/ml total activity.

Applicability At all times.

Action With the concentration of radioactive material released from the site exceeding Specification 3.17.1, immediately i restore concentration within the specification limits and '

report the event in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.2.3.

Bases This Specification is provided to ensure that the concentration of radioactive mater f als released in liquid waste effluents from the site to areas beyond the Site Boundary For Liquid Effluent (see Figure 5.1-4) will 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 I exposures within the limits of 10CFR 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 radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in

,h International Commission on Radiological Protection (ICRP)

Publication 2.

i 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 4

FACILITY CHANGE SAFETY ANALYSIS ~ LOG NO. 921 4 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 guides for the design objectives of 10CFR50 Appendix I. This is an incorrect assumption for the site specific environmental setting of Rancho-Seco,'therefore the Appendix I statement in the LCO has been deleted.

\

i

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

10. Existing Specification: l 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 dose 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 j 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 asoure that the .

releases of radict,ctive material in liquid effluents will be kept "as O low as reasonably achievable." The dose calculations in the ODCM f implement the requirements in Section III.A of Appendix I that {

conformance with the guides of Appendix I be shown by calculational j procedures based on models and data, such that the actual exposure of )

an individual through appropriate pathways is unlikely to 1 substantially underestimated. The equations specified in the ODCM 1 for calculation the doses due to the actual release rates of

. radioactive materials in liquid effluents are consistent with the l methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for l 1 1 ,

4 i

V l __ _o

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

10. Existing 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.

Applicability At all times.

1 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 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

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921' PROPOSED AMENDMENT No. 155 PAGE 37

10. New Specification (Cont.) !

required operating . flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I toL 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 erposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the-0DCM 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 exce'ss of the requirements of 40 CFR 141.

. Discussion:

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

l

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921-

- PROPOSED AMENDMENT NO. 155- PAGE 38

- 11. Existing 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 al1~ times Action With the quantity of radioactive material in any of the-listed tanks exceeding the above limit, immediately suspend 4

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.

i 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 i are two Regenerant Holdup Tanks. The limit applies to each tank ;

individually. 3 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

_ _ _ _ _ _ _ _ _ _ _ _ _ _ - - +

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

l

.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 ]

listed tanks exceeding the above limits, ininediately l 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. i Bases 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 l LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM.

Restricting the quantity of radioactive material contained in the specified outdoor tanks provides assurance that in i the event of an uncontrolled release of the tank's I contents, the resulting concentration at the nearest potable water supply and the nearest surface water supply d '

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

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 TREATMENT SYSTEM.

Discussion j The changes identify all the tank outfalls that will contribute to the Rancho Seco liquid ef7Tuent. 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.

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

l

12. Existing Specification:

N/A i

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 g 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 1 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.

0

. - - -____ - ____ ________ O

FACILITY GANGE SAFETY ANALYSIS LOG NO. 921 i PROPOSED AMENDMENT NO. 155 PAGE 41 f

12. New Specification: (Cont.)

requirement that the appropriate portions of this system be used wh=n 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. l Discussion:

This new Tech Spec is added to reflect the addition of LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM (a sluiceable demineralizar 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 dose design objectives set forth in Section.II.A of Appendix I of 10 CFR

50. )

\ 1 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 SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 42

13. Existing Specification:

3.18 GASEOUS EFFLUENTS 3.18.1 DOSE RATE The dose rate at and beyond the site boundary due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

a. The dose rate limit for noble gases shall be 300 mrem /yr to the total body and 3000 mrem /yr to the skin.

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

Applicability At all times

, Action g With the dose rate (s) exceeding the above limits, decrease R the release rate as soon as practicable to comply with the limit (s) given in Specification 3.18.1.

Bases 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, 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 specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For individual 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 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 j or to 3000 mrem /yr to the skin. These release rate limits also j restrict at all times the corresponding thyroid dose rate above j background to a child via the inhalation pathway to less than or i equal to 1500 mrem /yr. l

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

13. New Specification:

3.18 GASEOUS 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: l

a. The dose rate limit for noble gases shall be less than l 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, 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 /yr to

FACILITY CHANGE SAFETY 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 mrem /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 mrem /yr.

Discussion:

Adoption to the Standard RETS is also 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 relcased 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 i'

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 l with the above limits. .)

I Bases -l l

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 j guides set forth in Section IV.A of Appendix I to assure that the I releases of radioactive material in gaseous effluents will be kept i "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 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 I substantially underestimated. The dose calculations established in j 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 ftom 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

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/ , t FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT No. 155 PAGE 46 N

14. Existing Specification: (Cont.) \ /

Estimating Atmospheric Transport and Dispersion of Gaseous Effluents tj in Routine Releases from Light-Water-Cooled Reactors," Revision 1, ,'

l July 1977. The ODCM equations provided for determining the air doces '

! at or beyoyl the restricted area boundary (see Figure 3.18-1) vill he based on the historical average atmospheric conditions. P New Specification:

3.18.2 DOSE-NOBIF[ GASES 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.

1 Applicability At all times Action

a. With the calculated air dose from radioactive noble gases in gnaecus effluents exceeding any of the above i limits, prepare and submit to the Commission within 30 j days a Special Report pursuant to Specification 6.9.5. ThisjReport will identify the cause(s) for I 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 (a) to be taken to assure that subsequent releases w111 be in compliance with I the above annual limits.

Bases This specification is provided to implem rt 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 guidos set forth in Section II.B of Appendix 1. The ACTION statements provide the t

. l

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.c FACILITY CHANGE SAFETT ANALYSIS' LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 47 1

14. New Specification: (Cont.)

rj required operating flexibiliy/y.and at the same time implement the j

. !, M guices set forth in Section it.A of Appendix I to assure that the l

~-/ releases of radioactive material in gaseous effluents will be kept "as low as'is reasonably achievable,.' The Surveillance Requirement i imp 14 ment the requirements 1.n Section III.A of Appendix 1 that ]

cor.formance with the guides of Appendix I be shown by calculational '

procedures based on models and data such that the actual exposure of on individual through the appropriate pathways is unlikely to be substant.ially underestimated. The dose calculations established.in the ODCM for calculating the doses due to the actual release rates of {

radioactive noble gases in gaseous effluentu arn consistent with the i methodology provided in Regulatory Guide 1.109, " Calculation of 'f '

!- Annual Doses to Man from Routine Releases of Reactor Effluents for

!( the Pu yose 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 Diapersion of Gaseous Effluents in Routine Releases from Light-Wated Cooled Reactors," Revision 1, July 1977. The ODCM equatiotw provich.d'for determining that the air doses at the Site Boundary for Gaseous Effluents (Figure 5.1-3) are ,

based upon the historical average attv;Fpheric conditions.

Discussion, j 3.1 .

j This spec has been revised following Uta guidance in standard RETS. .

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

/

15. Existing Specification:

3.18.3 10 DINE-131, TRITIUM AND RALIC:"fCLIDES IN PARTICULATE FORM The dose or 648e commitner:pto a member of the public from I-131, from tritium. and fnom radionuclides in particulate form with half-lived 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 mree 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 frou with half-lives greater than eight days in gaseous effludnts 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 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 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 1 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. 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

1 l

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

15. Existing Specification: (Cont.)

due to the actual release rates of the subject materials are required 1 to be consistent with the Methodology provided in Regulatory Guide l 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 proride 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 exposure of man.

New Specification:

3.18.3 DOSE-IODINE-131,10 DINE-133, TRITIUM AND RADI0 ACTIVE 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 greater than eight days in gaseous effluent released to areas at or beyond the Site Bounda y 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 i

release of Iodine-131, Iodine-133, tritium, and radioactive l materials in particulate form with half-lives greater than

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMEt!DMENT NO.155 PAGE 50 i

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 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 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, "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 estimaring doses based l upon the historical average atmospheric conditions.

I 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, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption ,

I I

_ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ l

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

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FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENI*fENT NO 155 PAGE 52 1

16. Existing Specification: l I

3.19 GASEOUS RADWASTE TREATMENT 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 at 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 H 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 Radwaste Treatment System and/or Ventilation Exhaust Treatment System are not being used.

\

l 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 L 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 I 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 L--______-_-

FACILITT QIANGE SAFETY ANALYSIS LOG 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 Gaa 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.

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.

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

1 1

N_-__--_---__--_.-__--

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

16. New Specification: (Cont.)

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 environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix I to 10 CFR part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50.

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

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.

i.

FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 55

17. Existing Specification:

3.20 GAS STORAGE TANKS ,

1 The quantity of radioactivity contained in each wacte gas f decay tank shall be limited to 135,000 curies of noble j 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 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 erposure 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 l decay tank inventory of 98,414 C1 (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 p'reventing a further increase in gaseous activity.

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 FACILITY CHAN0E SAFE 1T 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 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 g 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 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 C1 (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.

4

1 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 j a PROCESS CONTROL PROGRAM to process wet radioactive wastes l to meet shipping and burial requirements.

Applicability At all times j Action f

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, waste principal chemical constituents, mixing and curing times.

New Specification:

3.21 SOLID RADIOACTIVE WASTES The solid radwaste systems shall be OPERABLE and used in j accordance with a PROCESS CONTROL PROGRAM for the {

SOLIDIFICATION and packaging of radioactive wastes to j ensure meeting the requirements of 10 CFR 20 and 10 CFR 71 prior to shipment of radioactive vastes from the site.

Applicability At all times Action j i

With the provisions of the PR.0 CESS CONTROL PROGRAM not i satisfied, suspend shipments of defective 1/ processed ,

or defectively packaged solid radioactive wastes from l I

the site.

- _ _ - i

4 FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921' PROPOSED AMENDMENT NO.155 PAGE 59

- 18. New Specification: (Cont.)

Bases The OPERABILITY of the solid radwaste system ensures that the system ,

will be available for use whenever radwastes require processing and. I 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, waste oil content, waste principal chemical constituents, mixing and curing times.

Di scussion'.

The action statement added the reporting requirement in accordance with !

the guidance in the Standard RETS. !

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 60

19. Existing 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 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 m41 function 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 meer .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 i condition shall be reported and described in the !

Annual Radiological Environmental Operating Report. i

1 l

1 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 j unavailable from any of the sample locations required I by Table 3.22-1, prepare and submit to the Commission 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 pathways.

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

^

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FACILITY 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. 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 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 report shall be submitted if: .

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

When radionuclides other than those in Table 3.22-2 -

are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of 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.

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 u_______.__ _

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l FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 63 1

19. New Specification: (Cont.) l 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 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 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.

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. 1 i

)

1

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 i 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 l 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. .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.

The detection capabilities required in Table 4.26-1 are ~

state-of-the-art for routine environmental measurements in industrial 1 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.

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.

I

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20. Existing Specification: Page 65

. Table 3.22-1

~ ~-

RADIOLOG bL ENVIRONMENTAL 50NITORING PR0' GRAM -- -

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

Frequency of Analysis

1. AIRBORNE A. Radiciodine 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 radioactivity greater- than or equal to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months -

following filter change. Perfors gamma isotopic

\ '

anal / sis on each sample where gross

beta activity is greater than 10 .i times the appro-priate control samples for the ~

same sample J period.

Perform gamma iso-

. topic analysis on composite (by location) sample

. at least once per ;

quarter. 1

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.

4

Sample locations are shown in the 00CM.

~

20. Existing Specification: (Cont.) Page 66

)

1

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Table 3.22-1 (Continued) i

, I RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM J Sampling and Exposure Pathway Number of Collection Type and Frequency and/or Sample. Samples

Frequency of Analysis

3. WATERBORNE

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 i quarter.

b. Runoff 1 Grab sample Gross Beta and ;

. coll (cted I-131 analysis of '

fortnightly. each suspended and

\ -

dissolved fraction.

Tritium analysis at

least once per quarter, plus gamma j isotopic analysis !

. on dissolved and suspended frac- -

tions.

c. Mud and 2 At least once Gross Beta on Silt smi-annually. each sample. l One pint sample '

of the top 3" of material 2 ft, from ,

shoreline.

Sample locations are shown in the 00CM.

(

20. Existinn Specification: (Cont.) Page 67

~

Table 3.22-1 (Continued)

, RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Sampling and Exposure Pathway Number of Collection Type and Frequency and/or Sample Sampl es* Frequency of Analysis

4. INGESTION

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

when animals are on pasture; at least once '

per month at other times.

b. Fish 1 At least semi- Gross Beta minus annually. One X 40 analysis on semple of each edible portion of of several each sample.**

species as

\ ^

shown in the 00CM.

i

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

each sampl e. **~ -

classes of food products as shown in the 00CM. ,

I

Sample locations are shown in the 00CM.

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

1 1

I

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 Sampl es* Frequency of Analysis

1. AIRBORNE A. Radioiodine 8 Continuous oper- Radioiodine canis-and Parti- ation of sampler ter. Analyze at culates with sample least once weekly collection as for I-131.

' 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 following filter change. Perform gamma isotopic g

analysis on each sample where gross beta activity is greater than 10 times the yearly 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 HANUAL.

9

- - --_-__~_________ _ _ _ _ _ _ __ _ _ _ _

20. New Specification: (Cont.) Page 69 n .

s 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 -

c 3. WATERBORNE .

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.

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 analysis of each sample. .

d. Mud and 2 At least once Gamma Isotopic ~

Silt semi-annually. analysis of Onepintsamgle 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.

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. 20. New Specification: .(Cont.) Page 70, Table 3.22-1 (Continued) i 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. I per month at d 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 j pie of each of edible portion 1 the several of each sample.

classes of food products as :/

shown in the REMP MANUAL.

Technical Specification sample locations are identified in the REMP MANUAL.

. I I

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

20. New Specification: (Cont.)

,- Discussion The changes made in Table 3.22-1 support additional sampling and collection frequency of the liquid effluent pathway documented in the REMP l

manual. The NRC considered that the District was deficiert 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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21.. Existing-Specification: Page 72 i .

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, i FACILITY CHANGE SAFETY ANALYSIS LOG.No. 921 PROPOSED AMENDMENT NO. 155 PAGE 74

21. New Specification:-(Cont.)

Discussion Table 3.22-2 vaines represent a reporting level of radioactive concentrations in environmental samples and not a regulatory limit. The requirements are a report that evaluates the effluent observations against i the EPA regulations documented in 40 CFR 190 (Tech Spec 3.25).

Additional reporting levels are included for cesium and iodine (I-131) analysis in fish and food products which are in the liquid effluent pathway. This is in conformance with the Standard RETS.

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,! FACILITY CHANGE SAFEIT ANALYSIS LOG No. 921 PAGE 75 r

$ PROPOSED AMENDMENT NO. 155

22. Existing Specification:

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

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

Aypicability At all times Action

a. With a land use census identifying a location (s) which yields a calculated dose or dere 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. W$th a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the 1 same exposure pathway) 20% greater than at a location from which samples are currently being obtained in B

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 done commitment (s) (via the same exposure pathway) may be deleted from this l' 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 the new location (s). 3

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 This specification is provided to ensure that changes in the use of I areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the i 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 1

l

i-FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921

. PROPOSED AMENDMENT NO. 155 PAGE 76

22. Existing Specification 1 (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 minimum garden size, the following assumptions were used: (1) that 20% of the garden was used for growing bread leaf vegetation (i.e.,

similar to lettuce rnd 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 annus11y 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 I 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 useo 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 calenlated 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 frem which samples are currently being chtained in accordance with Specification 3.21, add the new location (s) to the Radiological Environmental Monitoring Program within 30 days or submit a Special

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

22. New Specification: (Cont.)

Report to the Commissica 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 the 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 Pregram and the ODCM are made if required ~by the results of this census. This census satisfies the requirements of Section IV.B.3 or J.ppendix 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 j assumptions were used: (1) that 20 percent of the garden was used for growing broad leaf vegetatica (i.e., similar to lettuce and cabbage); and (2) a vegetation yield of 2 is/ square meter.

1 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 protection of the health and ,

safety of the general public. I 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

l m________ _ _ '

FACILITY CHANGE SAFETT ANAT3SD LOG NO. 921' PROPOSED AIGNDMENT No.155 PAGE 78 1

22. New Specification: (Cont.) !

monitoring program are mad' 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.

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- _ _ _ - _ - - - - - - i

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

23. Existing Specification:

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 i 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 releas'e(s) covered by this report. It shall also describe lecels 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 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 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 the plant radioactive effluents exceed twice the design objective doses of Appendix I. For the Rancho Seco site it is highly unlikely that the resultant dose to a MEMB!k JF THE PUBLIC will exceed the dose limits

I 1

1 FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 80

23. Existing Specification: (Cont.)

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 I PUBLIC to within the 40 CFR 190 limits. For the purposes of the I 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 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 do 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 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

23. New Specification: (Cont.)

releases to prevent recurrence of exceeding the above limitu and includes the schedule for achieving conformance with the above limits. This Special Repore, as defined in 10 CFR part 20.405(c), shall include ut 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 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 j 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 1 (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 conse utive 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 negligi' ale, 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 l violation of 40 CFR 190 have not already been corrected), in l

3 1

L___________ _ ___ )

I.

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 l PROPOSED AMENDMENT NO. 155 PAGE 82 l l

. 23. New Specification: (Cont.)

l l

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

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

24. Existing Specification:

3.26 INTERIABORATORY 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 acpuracy,of the measurements of radioactive material in environmental samples are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

l 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 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 pursuant to Specification 6.9.2.2.

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

24.'New Specification: (Cont.)

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 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.

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25. Existing Specification: Page 85 Table 4.1-1 (t.ontinueos ,

i INSTRUMENT SURVEILLANCE RtOV!REaEhis Channel Oescriotion Check Test (411 e ra te Remarks

44. Reactor oui tuing urain accumulation tank level NA NA K

43. Incore neutron detectors M(1) t4A NA (1) Check functioning, including fur;ctioning of c ueuter read-

, out and/or recocer reacout.

44 a. Process and area radi-ation monitoring systern W ct Q

b. Containment Area Monitors W NA R

45. Emergency plant radiation In:,truments Ml l) NA R (1) Battery eneck

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

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

48. Auxiliary Feednater Start Circuit

a. Phase imbalance /Under-puwar RCP $ dA R

b. Low Main Feedwater ;

i Pressure ,4A .4 R

\ -

49. Pressurizer Water Level M NA R I

50. Auxiliary Feedwater Flow Rate M NA R

51. Reactor coulan'. System Suo-cooling Margin Honitor M NA R

52. EMeV Pu..er Position Indicator (Primary Detector) .4 NA R

53. EMOV Position Incicator I i (dactup Detector) .4 4A R T/C or Acoustic l 54 DIOV Block Valve Position Indicator N NA R Sb. $4rety Valve Position in-dicator (Primary Detector) M NA R T/C

56. Sarety Velve Position In- l ofcator (Backup Detector) {

J Acoustic n l NA R l

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u ---- _ l

l 25..New Specification: Page 86 Table 4.1 1-(Continued) 1HSTRUMENT SURVElt. LANCE REQUIREMENTS:

Test Calibrate Remarts Channet Description Chect 42.' Reactor But1 ding drain NA NA R j

accumulation tank level M(1) NA NA (1) Checkfunctioning functioning,o f ,

43. Incore neutron detectors including coguter readout and/or recorder readout.

44. a. Process radiation R mor.itoring systes W Q

~

b. Area radiation

' monitoring system W H Q W NA R

c. Contairveent Area Monitors

45. Emergency plant radiation (1) Battery check M(1) NA R Instruments M(1) NA R (1) Check functioning

46. Envirmnaental air monitors NA R (1) Battery check

47. Strong motion accelerometer Q(1)

48. Aux 111ary Feedwater Start Circuit

a. Phase imbalance /Under ' '

S NA A Power RCP

b. Low Main Feedwater \ ,

Pressure NA M R N NA R l i 49. Pressurizer Water Level -

50. Auxiliary Feedwater Flow !

Rate M NA R 1

51. Reactor Coolant Systea -

R -!

Subcooling Margin Monitor M NA 52 OM0Y Power Position Indicator 4 (Primary Detector) H HA R )

l j

53. EMOV Position Indicator *

)

(Backup Detector)

T/C or Acoustic M NA R j i

j

54. EMOV Block Yalve Position 1 Indicator M NA R

55. Safety Yalve Position j l

Indicator (Primary Detector)

H NA R T/C

56. Safety Yalve Position '

Indicator (Backup Detector)

Acoustic M NA R Discussion: 1 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

26. Existing Specification:

4.19 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION Surveillance Requirements The maximum setpoint shall be determined in accordance with procedures as described in the.0DCM 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 materials in liquid effluents during actual releases. The alarm / trip set)oints for these instruments shall be calculated in accordance wit met 1ods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

New Specification:

4.19 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Surveillance Requirements The maximum setpoint shall be determined in accordance with methodology as described in the Offsite Dose Calculation Manual (00CM) 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 SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO.155 PAGE 88

26. New Specification: (Cont.)

i Records shall be maintained in accordance with the Process Standards A of all radioactive liquid effluent monitoring instrumentation M 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 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.

27.~ Existing Specification: -

Page 89

' Table 4.19-1 RADIOACTIVE 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) 0 NA R NA

a. Waste Water Flow

'" 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:

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

i

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 inuication 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.

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

1 h - - _ - - - - - - _ _ _ , _ _ _ _ _ , _ , _ __

27. New Specification: Page 90 Table 4.19-1 RADI0 ACTIVE LIQUID EFFLUENTiiONITORING INSTRUMENTATION SURVEILLANCE REQUIRCiENTS Instrument Instrument Channel Source Channel Channel Instrument Check Check Calibration Test

1. Gross Radioactivity Monitors Providing Alarm and Automatic Isol ation

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

Effluent Discharge M onitor

2. Flow Monitors

a. Regenerant Hold-up D(3) NA R Q Tank Discharge Line Total Flow

b. Waste Water Flow DI3) NA R Q Rate and Totalizer /IL w

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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMI!NDMENT NO.155 PAGE 91

27. New Specification: (Cont.)

Discussion New monitors are added on the Retention Basin Effluent, which is now the current Rand.a Seco environmental control point. The additions to the table notation are pursuant with Standard RETS.

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28. Existing Specification: Page 92 4.20 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING 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 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 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 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 requirements and General Design Criteria 60, 63, ana 64 of Appendix A .

to 10 CFR Part 50. l 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. :l The Reactor Building Purge Vent has two different flow rates, winter and 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 ano 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 addition transported to the manufacturer for calibration, the frequencies have been set as shown in Table 4.20-1.

l

28.-New Specification: Page 93 4.20- RADI0 ACTIVE GASEOUS 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 perfonnance of the INSTRUMENT CHANNEL CHECK, SOURCE CHECK, INSTRUMENT CHANNEL CALIBRATION, AND CHANNEL TEST at the frequencies shown in Table 4.20-1.

Records shall be maintained in accordance with the Process Standards of all radioactive gaseous effluent monitoring instrumentation alarm / trip-setpoints.

d Maximum setpoints and setpoint calculations shall be available for review to ensure 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 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 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 a constant release rate. However, releases from the Reactor Building may be at three different flowrates, winter, summer or minipurge.

Administrative controls assure that the correct flowrate is used.

The flow rate measurement devices are used only as flow indicating devices and &

not for actual measurement of flow rate. The actual flow rate of the M ventilation systems are periodically determined by surveillance procedures.

The flow rate devices must be removed from the ventilation system for the channel test, and in addition transported to the manufacturer 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 i Vent" names has been editorially changed to " Auxiliary Guilding Grade Level v .- 1 I

l

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29. . Existinst Specification: Pate 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 g(3)-

Activity Monitor Dill M Q(2)

NA NA

b. Iodine Sampler W NA

c. Particulate NA Sampler ~ W NA NA

d. System Effluent i 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)

Iodine Sampler NA NA

b. W NA

c. Particulate NA NA NA Sampler W

d. System Effluent l

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.-Existina Specification: (Cont.)' Page 95 Table 4.20-1 (Continued)

Instrument Instrument Channel Source Channel Channel Check Check Calibration Test Instrument I 3. Radwaste Service Area * ' ,

a. Noble Gas g(4)

' Activity Monitor DI1) 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.

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 perfomed using one or more reference standards.

(3) The Channel Test shall also demonstrate that automatic termination of this athway and control room alarm annunciation occurs if any of the {

fo11 ing conditions exist: -

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.

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29. Existing Specificar. ion: (Cont.) Page 96 .

. Table 4.20-1 (continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4 SURVEILLANCE REQUIREMENTS (4) The Channel Test shall also demonstrate that control room alann annunciation occurs if any of the following conditions exist:

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

b. Circuit failure.

c. Instrument indicates a downscale failure.

4 d. Instrument controls not set in operate mode.

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29. Ney Specification: Page 97 Table 4.20.-1 e

RADIOACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANGL REQUIREMENTS , , ,

^ ' ' '

~

. Instrument -

Instrument-Channel Source Channs1 Chinnel Instrument Check Check Calibration Test

1. ' Reactor Building Purge Vent

a. Noble Gas i Activity Monitor D M(4) R(3) ,QIII I 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 ,

l Flow Rate \

Measurement Device D NA R Q 4

2. Auxiliary Building Stack ,

i

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

b. Iodine Sampler W NA NA NA

]

c. Particulate Sampler W NA NA NA

d. System Effluent j 1

Flow Rate Device D NA R Q(6)

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

_____-t_-__ _ a

. , _ . .m._

- 29. New Specification: L(Cont.)i Page 98

- Tabl e . 4. 20-11 (Continued)

Instrument Instrument-Channel Source Channel. Channel Instrument Check Check Calibration Test

3. Auxiliary Building ,

Grade Level Vent

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

d. System Effluent D NA R -Q Flow Rate Device

e. Sampler Monitor D NA R Q.

Flow Rate l

Measurement

\

Device i

~

+

__.--_m _ - _ - _ . _ _ _ _ _ . _ _ _ _ _ _ _ ,

29. New Specification: (Cont.) .

Page 99-Table 4.20-1 (Continued) * ,

TABLE NOTATION (1) The CHANNEL. TEST shall also demonstrate that adtomatic 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- . .

J(2) The CHANNEL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exis'ts:

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

2. Circuit failure. .

3. Instrument indicates a downscale failure.

4. Instrument\ controls not set in operate mode.

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

(4) A check shall be performed prior to each relea'se.

(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 cc wot pose a personnel safety hazard (i.e., potentiel main steam safety actuation).

(7) The CHANNEL TEST shall also demonstrate that the 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 alarta/ 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.

I R0PO AMEND PA E lb0 l

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 l automatic termination of liquid releases, as listed in Table j 3.15-1, shall be used to limit the concentration of radioactive '

material released at any time from the site to areas beyond the I site boundary tr ne values given in Specification 3.17.1.

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 results of pre-release analyses shall be used with the calculational methods in the ODCM to\ assure that the concentration at the point of release is limited to the values in Specification 3.1'7.1.

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 ir exposures within: (1) the Section II.A Design Objectives of Appendix I,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 u y n 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 Protection (ICRP) Publication 2.

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO.155 PAGE 101

30. Existing' Specification: (Cont.)

There are no continuous releases of radioactive material in liquid effluents from the plant. All releases from 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 3.15-1. l The liquid effluent continuous monitor having provisions for automatic termination of liquid releases, as listed in Table 3.15-i, 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 Datch of liquid effluent to be 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 0FFSITE DOSE CALCULATION MANUAL (0DCM) to assure that the concentration at the point of release is limited to the limits of Specification 3.17.1. l 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 radioisotope and its MPC in air (submersion) was converted to any equivalent concentration in water using the methods described in International Commission on g

Radiological Protection (ICRP) Publication 2.

FACILITY CRANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMEIDMENT 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 objective's 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 plant 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 complianceLis removed from the bases statement in surveillance standards.

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31. Existing Specification: Page 103 Table 4.21-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Sampiing Minimus Analysis Type of Activity Lower Limit Liquid Release

Frequency Frequency Analysis Of Detectic Type (LLD)

(uCf/ml)(a)

Each Batch Each Batch ' Mn-54, Fa-59, 5 x 10-7 A. Batch Wasteb)(Re-lease Tanks Co-58, Co-60 P P In-65, Mo-99, Cs-134. Cs-137 Ce-141, and Ce-144 (c) 1-131 1 x 10-6

\

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

'i Entrained Gases (Gamma Emitters)

M Each Batch Composite (d) H-3 1 x 10'-5 .

P Gross Alpha 1 x 10-7 4 (d) 5 x 10-8 Each Batch Composi te Sr-89, Sr-90 P

)

y

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

RADIOACTIVE LIQUID WASTE S#4PLIhG AND ANALYSIS PROGRAN Table Notation a.- The lower limit of detection (LLD) is defined in the 00CM.

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 rep (cted. 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.

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

(uCi/ml)(a)

Each Batch Mn54 1E-7 A. Batch Waste lease Tanks b,d(Re-) Each Batch Fe59 4E-7 P P CoS8 SE-7 Co60 1E-7 Zn65 SE-7 Mo99 SE-7 80140 4E-8 Cs134 4E-8 Cs136 4E-8 Cs137 5E-8 Cel41 SE-7 Cel44 SE-7 (c) 1-131 1 x 10-8 g Dissolved and 1 x 10-5 Entrained Gases (Gamma Emitters)

H-3 1 x 10-5 l

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31.. New Specification: (Cont,) Page 106 i

TABLE 4.21-1 (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 3.17.1 (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 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 = 2.71 + 3.29(Br)0.5 3.70E4(YEYT) factor to account for Poisson distribution at very low where 2.71'c=ount rates.

background When background is estimated from a blank which has been counted for 9 a specific period, the following applies:

B = estimated background (counts) t

=b s tb h b = blank background (counts) tb = blank count time (seconds) ts = sample count time (seconds) r=1+ s tb i

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. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ b

31. New Specification: (Cont.) Page 106a i

TABLE 4.21-1 (Continued)

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation 3.70E4 w di si nteg ra ti o n s /sec o n d/mi c rocu ri e Y = yield of radiochemical process, i.e., the product of all factors such as emission fraction, chemical yield, qc.

E = counting efficiency (count / disintegrations)

V si sample volume (milliliters)

T = [1-exp(-Atc)]exp(-Atc_)_

A where x = decay constant (seconds -1) tc = time from midpoint of collection to start of counting (3) When spectroscopy is used to analyze the sample, the following LLD equation ig used:

LLD = 2.71 + 4.65(B)0.5 h, 3.70E4 (YEVT)

Where B is the counts in the Region Of Interest.

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31.-New Specification: (Cont.)s Page 106b-

. TABLE 4.21-1 -(Continued)

RADI0 ACTIVE-LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation

b. A batch release is the discharge of liquid wastes of discrete volume from the ' A' or 'B' Regenerant Holdup Tanks. Prior to sampling, each batch will be isolated, and then thoroughly mixed...to assure representative h

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 enalysis shall be reported as "less than" the instrument's LLD, and shall not be reported as being present. The "less than" values shal1~ 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 l 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%. These LLD's ensure compliance with 10 CFR 20 and also provide reasonable assurance of compliance with 10

'CFR 50, Appendix I dose guidelines. (Refer to Attachment B for added detail on LLD changes). Composite analysis for 10 CFR 50, Appendix I compliance has been moved to a new Table 4.21-2. Added conservatism was also included by increasing the sampling and minimum analysis of dissolved and entrained gases (gamma emitters),

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i 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 reasonabfy 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 i 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 Releases of Reactor Effluents for i the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix !

I," Revision 1, October 1977, and Regu12 tory 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:

4.21.2 Doses 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 Tables 4.21-1 and 4.21-2 and the methodology described in the Offsite Dose Calculation Manual (0DCM) g at the following frequencies:

a. Prior to the initiation of a release of liquid RADI0 ACTIVE EFFLUENT and, upon completion of each release a dose calculation update shall be made; and, [

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

32. New Specification (Cont.)

b.- Monthly, based on gamma-emitter and tritium analyses of RADI0 ACTIVE EFFLUENT releases during the previous calendar month i '

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 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 4 This specification is provided to implement the requirements of Sections II. A, III.A, and IV. A of Appendix 1,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth is :

Section II. A of Appendix I. The ACTION stacements provide the required operating flexibility and, at the same time, implement the guides set forth in Section IV.A of Appendix I-which assures, by definition, 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 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.

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 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 cf Implementing Appendix I,"

April 1977.

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

l FACILITY CHANGE SAFETY ANALYSIS LOG N0. 921 PROPOSED AMENDMENT N0.155 PAGE 110 l

\

32. New Specification (Cont.)

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

l The activity analysis of Cs-134, Cs-137 and I-131 at the Lower Limits j 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 provide an adequate basis for i determining the presence or absence of dose due to other radionuclides in plant liquid effluents, when no other indications are revealed during sample analysis.

1 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 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 excvss of the limits provided to implement 10 CFR 50, Appendix I, and 40 CFl.190.

The details of the dose tracking surveillance and new dose methodology is described in the ODCM. The addition of the LLD values in Tables 4.21-1 and 4.21-2 are specified in Attachment B of this safety evaluation.

Historical data was used in determining the LLD values and indicatbd that Cs-134, Cs-137, I-131 and tritium could be used as indicators for all nuclides in the Rancho Seco effluent mix.

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i 32. New Specification: (Cont.) Page 111 Tatile 4.21-2 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM bampling Minimum Analysis lype of Activity Lowr Limit Liquid Release Frequency Frequency Analysis (c) Of Detection Type (LLD)

(uCi/ml)(a)

Each Batch Composite (d) H3 1 X 10-5 A. Batch Waste lease Tanks b$ (R - P M Cr51 3 x 10-5 Mn54 1 x 10-7l Fe59 4 x 10-7i CoS8 5x10-7!

Co60 1 x 10-7:

Zn65 5 x 10-7!

Sr89 3 x 10-8' Sr90 3 x 10-8j 71 Zr95 Nb95 6 1 xx 10 10-8I Ag110m 1 x 10-7I

I131 1 x 10-8 Cs134

Cs136 9 x 10-9 8

h Cs137 4 x 10 1 x 10- 8

Ba140 4 x 10-8 Ce141 5 x 10-7 Ce144 5 x 10-7 Gross Alpha

1x10-7l I

i

For these short-lived isotopes, the LLD specified in Tables 4.21-1 and 4.21-2 are met during pre-release analysis, and as such the LLD does not have to be met for the composite sample analysis.

32. New Specification: (Cont.) Page 112 TABLE 4.21-2 (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 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 = 2.71 + 3.29(Br)0.5 b 3.70E4(YEVT) where 2.71 = factor to account for Poisson distribution at very low backgroundhount rates.

3 When background is estimated from a blank which has been counted for a specific period, the following applies:

B = estimated background (counts) tb b = blank background (counts) tb = blank count time (seconds) ts = sample count time (seconds) r=1+t s .

tb l

V32. 'New Specification: .-(Cont . ) .Page 112a TABLE 4.21-2 (Continued)-

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation 3.70E4 = disintegrations /second/ microcurie i

Y: = yield of' radiochemical process, i.e., the product of all factors .j such as emission fraction, chemical yield, etc. q E =' counting efficiency (count / disintegrations)

Y = sample volume (milliliters)

T=[1-exp(-Ats)]exp(-Atel A

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

(3) When spectroscopy is used.to analyze the sample, the following LLD equation is used:

LLD = 2)71+4.65(B)0.5 3.70E4 (YEVT)

Where B .is the counts in the Region Of Interest.

s

32. New Specification: (Cont.)- Page ll2b

-t

.,4 m 4 f l l

I< TABLE 4.21-2_ (Continued)

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM

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

the A or B Regenerant Holdup Tanks. 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 t

than" values shall not be used in the ODCM evaluations. ~ However, if the

d. -

nuclide is measured and identified at a value less than the Table 4.21-2 LLD_ value, it shall be reported and entered into the ODCM. evaluations.

1

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.

Discussion:

A definition of LLD has been added to the Table notation of Table 4.21-2 j as detailed in Attachment 3 of this Safety Analysis. j k

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1 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. l li New Specification: .

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4.21.3 LIQUID HOLDUP TANKS * .f l

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 i 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 I 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 i limits of 10 CFR Part 20, Appendir B, Table II, Column 2.

1

Tanks included in this specification are those outdoor tanks that I are not surrounded by liners, dikes, or walls capable of holding the l 1

o FACILITY CHANGE SAFETY 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.

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j 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 days in accordance with the methodology and parameters in the 0FFSITE DOSE CALCULATION MANUAL (ODCM) when LIQUID j EFFLUENT RADWASTE TREATMENT SYSTEMS are not being fully utilized.

The installed LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM shall be I' considered OPERABLE by meeting Specifications 3.17.1 and 3.17.2.

Bases The OPERABILITY of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM I 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 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 Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as the dose design objectives set forth in Section II.A of <

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

1 1

The installation of the LIQUID EFFLUENT RADWASTE TREATMENT SYSTEM is i not complete. This specification will become ef fective when the {

system is declared operable. !

l Discussion l l

The new surveillance requirement is added per District commitment to the /

NRC, assuring the operability of the radwaste treatment system in the )

event of liquid effluent release. l l

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35. Existing Specification: Page 116 1

4.22. GASEOUS EFFLUENTS !

4.22.1 Dose Rate Surveillance Requirements i The release rate of noble gases in gaseous effluents shall be controlled by 7he offsite dose rate as established'in Specification 3.18.1. The dose rate -

shall be detennined in accordance wit,h the 00CM.

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 detennined by obtaining representative samples and performing 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 detennined to be within the required

- limits by using the results of the sampling and analysis program, specified in

( Table 4.22-1, in perfdnning the ca+culations of dose rate beyond the site boundary in accordance with the ODCM.

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 reasorable 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 Appendix 8, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase 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 / 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.

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35. New Specification: Page 117 i L

4.22 GASE0US EFFLUENTS 4.22.1 Dose Rate Surveillance Requirements 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 methodology describcd in the 0FFSITE DOSE CALCULATION MANUAL (0DCM).

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 relesse 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, specf fied 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 accordance 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 effluer.ts 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 i 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 i 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 or equal to 1,500 mrem / year for the nearest dairy cow to the plant.

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36. Existing Specification: Page 118 Table 4.22-1 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM 1

Sampiing Minimum Analysis Type of Activity Lower Limit ,

Gaseous Release Frequency Frequency Analysi s of Detection I i

l Type - (LLD) a (uCi/ml)

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

Storage Tank Grab P Emitters (f) l Sample j l

P

.]

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

Sample (c) p P H-3 1 x 10-6 i

C. Auxiliary M'(b, c ) M(hr) Principal Gamma 1 x 10-4 ll Building Stack, Grab Emitters (f) s' and Radwaste Sample .

Service Area H-3 1 x 10-8 l Vent D. All Release Continuous W(d) I-131 1 x 10-12 Types as listed Charcoal in A,B,C above Sample Continuous W(d)

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

(I-131, Others)

Continuous M Gross Alpna 1 x lu-"

Composite Particulate Sampie Continuous Q Sr-89, S r-90 1 x 10-11 Composite Particulate Sampie Continuous Noble Gas Noble Gases 1 x 10-4 Monitor Beta or Gamma as Xe-133 (Gross) ,

- L

A

36. Existing Specification: (Cont.) , Page 119.

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Table 4.22-1 (Continued)

RADIOACTIVE GASEOUS WASTE SAMPLING AND AMALYSIS PROGRAM

~

Table Notation -

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

b. Analysis shall also be perfomed when gross beta-gamma activity analysis of reactor coolant indicates greater than 10 uCi/mi and after each 10 pCi/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 j source. l

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 perfomed when reactor coolant indicates 10uct/ml gross beta garna activity and every 10nci/ml increases thereafter. 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 V of 10.

p e. Tritium grab samples shall be taken at least daily during refueling activi ties.

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 also be

< reported.

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36. N w So-cifiention: .. Pags 120 r.H..: ,is,r.;. nyLG u s ru v.ed RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards i

Table 4.22-1 RADI0 ACTIVE GASEOUS WASTE SAMPLING 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 Gamma 1 x 10-4 Grab Emitters (f)

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

Sample (b,e,1)

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

N H-3 1 x 10-6 D. Auxiliary M('b) M(b) Principal Gamma 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 in A,B,C,0 above Sample I-133 1 x 10-10 Continuous W(d)

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

(I-131, Others) ..

Continuous N Gross Alpha (h) 1 x 10- "

Composite Particulate Sample Sr-89, Sr-90(g) 1 x 10-11 Continuous Noble Gas Noble Gases 1 x 10-6 Moni tor Gross Beta and Gamma as Xe-133

36. New Specification: (Cont.) Page 121 Table 4.22-1 (Continued)

RADI0 ACTIVE GASE0US 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 a per unit volume, which is required to be detected, if present, in G 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 LLD in microcuries per cubic centimeter is' given by the equation:

LLD = 2.71 + 3.29(Br)0.5 3.70E4(YEVT) where 2.71 = factor to account for Poisson distribution at very low background count rates.

When backgIound is estimated from a blank which has been counted for a specific period, the following applies:

B B = estimated background (counts) t

=b s th b = blank background (counts) tb = blank count time (seconds) ts = sample count time (seconds) r=1+ s Ib l

36. New Specification: (Cont.) Page 12'la Table 4.22-1 (Continued)

RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation 3.70E4 = disintegrations /second/ microcurie !b Y = yield of radiochemical process, i.e., the product of all factors such as emission fraction, chemical yield, etc.

E = counting efficiency (count / disintegrations)

V = sample volume (cubic centimeters) .

where 1 = decay constant (seconds -1) tc = time from midpoint of collection to start of counting (3) When spectroscopy is used to analyze the sample, the following LLD equation is used:

LLD = 2.71 + 4.65(B)0.5 g 3.70E4 (YEVT)

Where B is the counts in the Region Of Interest.

b. Analysis shall glso be performed when gross beta or gamma activity anaylsis of readtor coolant indicates greater than 10 pCi/ml and af ter each 10 pCi/mi increase in the gross beta or gamma activity analysis.

4 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.

h t

36. New Specification: (Cont.) Page 121b I

Table 4.22-1 (Continued) '

RADIOACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PROGRAM Table Notation ]

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 perfomed when reactor coolant indicates 10pC1/ml gross beta gamma activity and every l 10pCi/ml increases thereafter. 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 maybe increased by a factor of 10.

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, Xb133, Xe-133m, Xe-135, Xe-135m, and Xe-138 for gaseous samples and Mn-54, Fe-59, Co-58, Co-60, Zn-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 6.9.2.3. Nuclides which are below the LLD for the analysis shall be reported as "less than" the nuclide's LLD and shall not be reported as being present at the LLD level for that nuclide. 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 pCf /cc gross beta activity then a Sr-89, Sr-90 analysis will be performed on those samples exceeding this value.

h. Gross alpha performed on a monthly basis for each environmental release particulate sample. This fulfills the requirements of perfoming a monthly composite, h

i. Af ter purging seven reactor building volumes, a technical evaluation, prior to reinitiation of a purge following an out of service period, may be conducted in lieu of sampling and analysis.

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i FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PEOPOSED AMENDMENT NO. 155 PAGE 122

37. Existing Specification: J 4.22.2 N0BLE GASES l Dose Calculations Cumulative air dose contribution for the quarterly or yearly period ,

as applicable shall be determined in accordance with the Offsite Dose l 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 scatements 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 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 Compliatee 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 will be based upon the historical average atmospheric conditions. NUREG-0133 provides methods for dose 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 calendar year shall be determined in accordance with the methodology described in the OFFSITE DOSE CALCULATIONAL MANUAL (0DCM) at least monthly.

FACILITY CHANGE SAFEIT ANALYSIS LOG NO. 921 PROP 0 SED AMENDMENT NO. 155 PAGE 123

37. New Specification: (Cont.)

l l 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 ret forth in Section II.B of Appendix I. The ACTION statements provide the '

l required operating flexibility and at the same time implement the goides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance 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 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 Easeous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Dobes 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 Routice 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. i i

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38. Existing Specification: Page 124 4.22.3 Iodine-131, Tritium and Radionuclides in Particulate Form Dose Calculations Cumulative dose contributions for the quarterly or yearly period as applicable shall be detennined in accordance with the ODCM at least monthly. ,

1

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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 ga'seous 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 consi t tent with the methodology provided in b Regulatory Guide 1.109, '" Calculating of Annual Doses to Han from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 ano Kegulatory Guide 1.111, 11 l

" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous l Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, (

July 1977. These equations also provide for detennining the actual doses I based upon the historical average atmospheric conditions. The release rate I specifications for radioiodines, radioactive material in particulate fonn and radionuclides other than noble gases are dependent on the existing j radionuclides pathways to man, beyond the site boundary. The pathways which l are examined in the development of these calculations are: (1) indivioual l inhalation of airborne radionuclides, (2) deposition of radionuclides onto l f

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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38. Nw Specification: Page 125

(

l l 4.22.3 Dose-Iodine-131, Iodine-133, Tritium, and Radioactive Materials in Pa rticulate Form.

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.

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 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 substantially underestimated. The ODCM calculational methods for calculating the doses due to thg 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 y for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," 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," Revf sion 1, July 1977 These equations (

also provide for estimating doses based upon tha historical average _l atmospheric conditions. The release rate specifications for radioiodines, and ,

radioactive material in particulate form are dependent on the existing j radionuclides pathways to man at or beyond the Site Boundary for Gaseous l Ef fluents -(Figure 5.1-3). The pathways which are examined in the development j of these calculations are: (1) individual inhalation of airborne I radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by en, (3) deposition onto grassy areas where ,

milk animals and meat producir,g animals graze with consumption of tha milk and meat by man, and (4) deposition on the ground with subsequent exposure of man. l l

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l FACILITT CHANGE SAFETY ANALYSIS' LOG NO. 921-

PROPOSED AMENDMENT NO. 155 PAGE 126

39. , Existing Specification:

l 4.23 GASEOUS RADWASTE TREATMENT l >

Dose Proiagtions 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 cf the guide set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous

,, effluents.

New' Specification: .

4.22.4 GASEOUS RADWASTE TREATMENT Surveillance Requirement .

I Doses due to gaseous releases to areas at and beyond the Site !

Botadary For Gaseous Effluento (see Figure 5.1-3) shall be projected ,

at least once per 31 days in accordance with the methodology and l parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM) when Gaseous Radwaste Treatment Systems are not being fully utilized.

The installed VENTILATION EXHAUST TREATMENT 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

FACILITY CHANGE SAFEIT ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 127

39. New Specification: (Cont.)

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 50.36, 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. l

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l 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 l daily when radioactive materials are being added to the tank and the Reactor Coolant System activity exceeds the limits of Specification 3.1.4.

l 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 I 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 i' 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.

_. ____._m _.______________.____..________________w

h- FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT NO. 155 PAGE 129

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 patch of the same type of wet waste until at least 3 causecutive 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. .I Reports i

The semiannual Radioactive Effluent Release Report shall include the following Ij information for each type of solid waste shipped offsite during the report i period: l 1

a. Container volume,

b. Total curie quantity (determined by measurement or estimate),

c. Principal radionuclides (determined by measurement or estimate),

d'. Type of wa'ste (e.g., spent resin, compacted dry waste evaporator bottoms),

e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and

f. Solidification agent (e.g., cement, urea formaldehyde).

9 g e 4

41. Existing Specification: (Cont.)' -

Pa8e 131 1

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 *he 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.

4 1

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41. New Specification: Page 132 N

4.25 SOLID RADIOACTIVE WASTES ;

~ Surveillance Requirements

' 4.25 T - The solid radwaste systems shall be demonstrated OPERABLE at least

~ once per 92 days by:

a. ' Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRM, or 4

b. Verification of the existence of a valid contract for o~ SOLIDIFICATION to be perfor1ned by a contractor in accordance with a PROCESS CONTROL PROGRM.

4.25.2 The PROCESS CONTROL PROGRM shall be used to verify the-SOLIDIFICATION of at lease one representative test specimen from at l 1 east 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 PROGRM, and a ' subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters ~

determined by the PROCESS CONTROL PROGRM.

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 sting of representative test specimens from each consecutive b itch of the same type of wet waste until at least' 3 consecut . initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRM 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 4 parameters included in establishing the PROCESS CONTROL PROGRM 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.

~

____________mu____ ___ _ a$ d

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.

t

-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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l FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 134

42. Existing Specification:

4.26 RADIOLOGICAL ENVIRONMENTAL MONITORING Surveillance Requirements The radiological environmental monitoring sampics 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 exposures 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 exposure pathways.

The specified monitoring program is in effect at the present time.

Program changes may be initiated based on operational erperience 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 I

Surveillance Requirements l 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 erposures 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 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

FACILITY CHANGE 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. Existing Specification: Page 136..

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43.:New Specification.. Page 137.

RANCHO $ECO UNIT 1 I TECHNICAL $ SPECIFICATIONS Table 4.26 1

' MAXIMUM YALUES FOR THE LOWER LIMITS OF DETECTION (LtD)a, d Milk Food Products Mud and $11t Water Airborne Particulate Fish (PCf/kg, wet).

(pCf/kg, wet) (PC1/1) (PC1/kg, wet)

Analysis (pCf/1)~ orGases(PC1/m3)

q gross beta 4(b) , 1 x 10 2 3rd 2000(1000(b)) ,

130

$4Mn : 15-30 , 260

$9Fe 130 150 l 58Co 15 150 130 60Co 15 260 65Zn 30 952 r.Nb '15(e) 1 60 131g 1 (b) '7 x'10-2 .

15 60 150 2'x 10 2(c) 130 134Cs 15 (10(b))

18 80 180 1 a 10-2(e) 150 137Cs is (10(b))

15(e) 14084.La 15 (e)

\

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).

1

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~ FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT No. 155 PAGE 138 i

'44. Existing Specification: l 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 i 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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L 44.. New Specification: (Cont.) Page 139 Table 4.26-1 (Continued) l MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a, d L Table Notatioa

a. (1) The lower limit of detection (LLD) for a radionuclides presented in this table is the largest concentration, expressed in picocuries per unit sample, which is required to be detected, if present, in order to achieve compliance with the applicable j g regulation, given stated operating conditions and calculation methology.

(2) The LLD of a radioanalysis system is that value which will l 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 picocuries per unit sample is given by the equation:

LLD = 2.71 + 3.29(Br)0.5 3.70E4(YEVT) h

( 2.71 '=\ factor to account for Poisson distribution at very low background count rates.

When background is estimated from a blank which has been counted for a specific period, the following applies:

B = estimated background (counts)

=b s tb h b = blank background (counts) tb = blank count time (seconds) ts = sample count time (seconds) r=1+ s tb

44. New Specification: (Cont ).

Page 139a Table 4.26-1. (Continued)

MAXIMUM VALUES FOR-THE LOWER LIMITS OF DETECTION (LLD)a,- d Table Notat! ion 3.70E4 = disintegrations /second/ microcurie Ib Y = yield of radiochemical process E = counting efficiency (count / disintegrations) l V = sample volume (liters) or mass (kilograms)

T = [1-exp(-Ate)]exp(-xtc)_

A.

where x = decay constant (seconds -1) tc = time from midpoint of collection to start of counting (3) When spectroscopy is used to analyze the sample, the following LLD equation is used:

LLD + 2.71' + 4.65(B)0.5 3 70E4 (YEVT)

Where B is the counts in the Region Of Interest. .

-___m___._-_._m_ __m__ . _

44. 'New Specification: (Cont'. ) Page 139b j i

i Table 4.26-l'(Continued)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION-(LLD)a,.d Table Notation !

(4) Analyses shall be performed in such a manner that the stated LLD's; will be achieved under routine conditions. Occasionally, background d

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 sgown ig for composite analysis.

For individual samples, 5x10 pCi/m3 is..the LLD.

d. Other peaks which are measurable and identifiable,- together with the nuclides in Table 4.26-1, sha?1 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 ake defined for Cesium detection in the 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).

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921

PROPOSED AMENDMINT NO. 155 PAGE 140

45. Existing Specification: I l

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.

Reports The results of the land use census shall be included in the Annual !

Radiological Environmental Operating Report. l

.\ '

Bases This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that j 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 l

\

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 cab.bage), and (2) a vegetation yield of 2 kg/ square meter.

New Specification:

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.

The land use census or portions thereof, shall be conducted during the appropriate time of the year to provide the best results. ;

I Reports i The results of the land use census shall be included in the Annual Radiological Environmental Operating Report.

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 made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored, since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy 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 vegetation 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.

l l

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46. Existing Specification: Page 142 l

. 4.29 FUEL CYCLE DOSE l

Surveillance Requirements Cummulative dose contributions from liqu.J 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 j 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. 1 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 assumed that the dose commitment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose j 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 pubite 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 determined 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 numerica.1 guides for design objective doses of Appendix 1 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 (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 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 The modifications are in conformance with the Standard RETS.

i l

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 i 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 l l

(1) FSAR paragraph 1.2.1 d) FSAR paragraph 2.2.1 New Specification:

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 I 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 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.

- .- ~ - _ ._ . - - - - _ _ _ _ - - _ _ _ _ . . _ - _ _ . -_______ _

. 'l J

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 '

PROPOSED AMENDMENT NO. 155 PAGE 145 i

i

47. New Specification: (Cont.)

5.1.4 Site Boundary for Liquid Effluents 1

I The SITE BOUNDARY FOR LIQUID EFFLUENTS shall be shown in Figure 5.1-4.

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i FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 150

47. New Spe.cification (Cont.)

i Discussion: l f i The addition of site schematic drawings identifying the affected areas of e ( l effluent release is pursuant with the Standard REIS. (

                                                                                                             \
                  -                     _ _ . _ - _  _m.______.______..__________________________         ___ _ _ _ _ _ _ _ _ _ _ _ _ _

I . FACILITT 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 l 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. l Safety Analysis For Technical Specification Changes (Non-Administrative / Editorial) Items 1 -7 7 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 cnd 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 Staud*rd Radiological Effluent Technical

  &                                                        Specifications (RETS) in NUREG-0472 and NUREG-0452 compliance with 10 CFR 50, Appendix 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) 12strumentation

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 secount the maximum undiluted liquid waste ;

flow, the combined liquid releases for each unit, and the mixing ! i effects in the receiving water body in the near field of the { !- discharge structure. For plants with non-recirculating main j condenscr cooling systems, the mixing effects in the receiving water J body are ignored for conservatism. However, for plants with recirculating cooling systems, where cooling water discharge flow i j l j

c v i I LOG No. 921 FACILITY CHANGE SAFETY ANALYSIS PROPOSED AMENDMENT NO. 155 PAGE 152 Safety Analysis For Technical Specification Changes (Non-Administrative / Editorial (Cont.)

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.

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 after 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 bacis 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 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 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 requirements (Tech Spec 4.21.1) for radionuclides concentrations in 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 had 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. Thu setting of the LLD's for Cs-134 and Cs-137 at a concentration

b I p b FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921

l. PROPOSED AMENDMENT NO.155 PAGE 153

1. Concentration (Cont.-)

equivalent of 50% of Technical Specification values has been I 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 (0DCM) will address dose related contributions from liquid effluents, thereby, addressing ALARA. The Offsite Dose Calculation Manual (0DCM) 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 l 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 significant 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 CFR 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 Tables 4.21-1 and 4.21-2 reflect 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

c - _ - - _ _ , 8 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 j than the thyroid. I-131 was found to contribute essentially 100 % of I the dose to thyroid. Tritium contributes variable fractions of the l total dose, but is considered separately due to the distinct analysis j method. j i I l l l

FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 PROPOSED AMENDMENT 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 Technical Specification 3.17 2 (assuming an estimated annual plant effluent out flow of 20 million gallons and an average dilution flow rate of 5,000 spm). 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 I 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; i l

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l l 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 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 10 CFR 50, Appendix I is maintained. 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 provide reasonable assurance of continued compliance with the ALARA dose guidelines of 10 CFR 50, Appendix I. The dose tracking system is required to be operated and maintained such that operating 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 projected to the end of the quarter and/or year, to exceed the ALARA dose guidelines of 10 CLR 50, Appendix I. The use and definition of the Maximum Hypothetical Individual is pursuant to 10 CFR 50, 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.

3. Radiological Environmental Monitoring Program The Rancho Seco Radiological Environmental Monitoring Program l 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 This program is based on the guidance of the model

(Table 3.22-1). program in NUREG-0472, Rev. 2, and NUREG-0452, Rev. 5 (draf t). 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 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. 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 erposure pathways that exist downrtream 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

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 j 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 leed 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 erposure pathways. This change does not . S 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 erposure 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. l

FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 158

4. Land Use Census The Rancho Seco REMP, utilizing the guidance of 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 I I

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 j 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 1 The proposed change to the Technical Specification regarding the Rancho , Seco site land use census does not involve a significant hazarda i consideration because operation of Rancho Seco in accordance with this l 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 i 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.

l k l FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMERDMENT NO. 155 PAGE 159

5. Reporting, Procedures And Audits (Cont.)

Basis For No Significant Hazards Determination The proposed changes to the Technical Specifications regarding the review, audit, procedures and reporting 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 incorporate administrative improvements and do not alter plant design or safety / 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 ofg 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. Ins trumenta tion Radioactive gaseous and liquid effluent monitoring instrumentation monitor: 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 (ODCM) 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 (draft) and NUREG-0472, Rev. 2, and to reflect current plant design and operation. ,

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FACILITY CHANGE SAFETY ANALYSIS LOG No. 921 , PROPOSED AMENDMENT NO. 155 PAGE 160

6. Instrumentation (Cont.)

The existing RSTS Table 3.15-1/4.19-1 specify operability and surveillance requirements for the Regenerant Holdup Tank (RHUT) Discharge Line Monitor. This monitor functions as the plant liquid effluent monitor, providing automatic termination 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 more closely conform with the standard RETS. Reasonable operability and surveillance requirements are established for the RHUT total flow monitor to ensure that the total volume'of water released from the A & B RHUT (effluent control point) to the Retention Basin is known for the determination of offsite dose. Additional changes 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. Radioactive gaseous effluent monitoring instrumentation surveillance frequencies for the instrument channel calibration in Table 4.20-1 are decreased from monthly to refueling pursuant to the guidance contained in NUREG-0452, Rev. 5 (draft). Notations are added to ensure that the\ channel test for the Auxiliary Building Stack noble gas activity monitor adequately demonstrates functionality. - Allowances is provided for not performing channel testing of the 4 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 proximity 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 operation 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; I

    ,y FACILIfi #40E SAFKlT ANALYSIS                                             LOG NO. 921 PROPOSED LeJ DMENT 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 potentie.1 radioactive liquid effluent streams in addition to the RHUT discharge flow path. The clarifications to the. alarm / trip setpoints easures that the limit of 10 CFR 20.106 are maintained. There are no changes to system functions and therefore l these changea do not create the possibility of a new or different I 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. t ,

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N FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 PROPOSED AMENDMENT NO. 155 PAGE 162 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 L not be created d'ue to these editorial and administrative changes. l 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 control relating to radiological effluent releases and will preserve 3 the margin of safety. Therefore, the administrative changes will not I > reduce the margin of safety. l

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v t e Letter Sequence Other
TAC:56033, (Open) TAC:64735, (Approved, Closed) TAC:64736, (Open)
Initiation Acceptance... Administration Meeting Questions Answers Results Approval... Other: ML20125B767, ML20127N492, ML20128Q064, ML20128Q654, ML20140G645, ML20147B176, ML20148G258, ML20148G286, ML20196B675, ML20196B690, ML20214M133, ML20235R405
MONTHYEAR1987-10-26 [Table View]

ML20235R405

Text

Dear Mr. Miraglia:

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