ML20196D909
| ML20196D909 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/11/1988 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20196D905 | List: |
| References | |
| NUDOCS 8802240043 | |
| Download: ML20196D909 (21) | |
Text
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f RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.11 FIRE SUPRESSION SYSTEMS 1.11.1 The FIRE SUPRESSION HATER SYSTEM shall consist of water sources, pumps and distribution piping with associated sectionalizing control of isolation valves. Such valves include yard hydrant valves and the first valve ahead of the water flow alarm device on each sprinkler header, hose standpipe or spray system riser which
.. protect nuclear safety components.
1.11.2 The FIRE SUPRESSION CARBON DIOXIDE SYSTEM shall consist of a CO2 source and distribution piping with sectionalizing control valves which protect nuclear safety components.
1.12 STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, trains or designated components obtained by dividing the specified test interval into n equal subintervals,
- b. The testing of one system, subsystem, train or designated components during each subinterval.
155- 1.13 PROCESS CONTROL PROGRAM PROCESS CONTROL PROGRAM (PCP) - The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION /DEHATERING of radioactive wastes from liquid systems is assured.
1.14 SfLIDIFICATION 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).
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l _ Proposed Amendment No. 155, Revision 2, Supplement 1 8802240043 880211 1-6 PDR ADOCK 05000312 DCD
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T FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921, REV. 2, SUPP. 1 PROPOSED AMENDHENT NO. 155, REV. 2, SUPP. 1 PAGE 2
- 1. Existina Soecific31100:
1.13 PROCESS CONTROL PROGRAM 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 Soecification:
1.13 PROCESS CONTROL PROGRA3 PRCCESS CONTROL PROGRAM (PCP) - The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION /DEHATERING of radioactive wastes from liquid systems is assured.
1.14 SOLIDIFICATION SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed),
monolithic, immobirized solid with definite volume and shape, boundad 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 Solidifice. tion.
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 155- 3.17.4 Liauid Effluent Radwaste Treatment The LIQUID EFFLUENT RADWASTE TREATHENT 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 when projected doses due to the liquid effluent beyond the Site Boundary For Liquid Effluents (see Figure 5.1-3) when averaged over 31 days, would exceed 0.25 mrem to the total body or 0.83 mrem to any organ.
Aeoli cabili ty At all times.
Action
- a. With the LIQUID EFFLUENT RADHASTE TREATHENT 3YSTEM inoperable for more than 31 days or with radioactive liquid waste being discharged without treatment and in excess of the above li.,its, prepare and submit to the Commission within 30 days pursuant +o Specification 6.9.5 a Special Report which includes the following information:
- 1. Explanation of why liquid radwaste was being distharged 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 EFFLUENT RADHASTE TREATHENT SYSTEM ensures that l this system will be available for use whenever liquid effluents require treatment prior to release to the envirorment. The requirement that the appropriate portions of this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in liquid effluents are maintained "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 l
l governing the use of appropriate portions of the LIQUID EFFLUENT RADHASTE l TREATHENT SYSTEH are the dose design objectives set forth in Section II.A of l Appendix I, 10 CFR Part 50, for liquid effluents. ,
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l Proposed Amendment No. 155, Revision 2, Supplement 1 3-72a
FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921, REV. 2, SUPP. 1 PROPOSED AMENDHENT NO. 155, REV. 2 PAGE 39
- 12. Existino Soecification:
N/A New Soecification:
3.17.4 LIOUID EFFLUENT RADWASTE TREATHENT The LIQUID EFFLUENT RADRASTE TREATHENT 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 when projected doses due to the liquid effluent beyond the Site Boundary for Liquid Effluents (see Figure 5.1-3), when averaged over 31 days, would exceed 0.25 mrem to the total body or 0.83 nrem to any organ.
Acolicability At all times Action
- a. Hith the LIQUID EFFLUENT RADWASTE TREATHENT SYSTEM inoperable for more than 31 days or with radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.5 a Special Report which includes the following information:
- 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
HAlai i
The OPERABILITY of the LIQUID RADHASTE TREATHENT SYSTEM ensures that this system will be available for use whenever liquid effluents require l
treatment prior to release to the environment. The requirement that
FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921, REV. 2, SUPP. ,1 PROPOSED AMENDHENT NO. 155, REV. 2 PAGE 40
- 12. New Soecification: (Cont.)
the appropriate portions of this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in liquid effluents are maintained "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 RADHASTE TREATHENT SYSTEM are the dose design objectives set forth in Section II.A of Appendix 1, 10 CFR Part 50, for liquid effluents.
Discussion:
This new Technical Specification is added to reflect the addition of LIQUID EFFLUENT RADWASTE TREATHENT SYSTEM (a sluiceable demineralizer 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 ;n Section II.A of Appendix I of 10 CFR 50.
Because Rancho Seco is a dry site, the Standard Technical Specification 1/4 factor values of 0.06 mrem to the total body and 0.2 mrem to any organ would require all A & B RHUT water be processed through the Sluicable Demineralizers. This requirement would be over restrictive and an unnecessary condition for operation. The values chosen, which represent one-twelfth (1/12) of the annual 10 CFR 50 Appendix I guidelines, will ensure ALARA and provide more flexible operations.
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 3.25 FUEL CYCLE DOSE 155- The dose or dose commitment to any real MEMBER OF THE PUBLIC due to releases of radioactive material in gaseous and liquid effluents and to direct radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is
- limited to less than or equal to 75 mrem) in a calendar year.
Aeolicability At all times Action 155- a. Hith the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.17.2.a. 3.17.2.b, 3.18.2.a. 3.18.2.b, 3.18.3.a, or 3.18.3.b, or exceeding the reporting levels of Table 3.22-2, calculations shall be made including direct radiation contributions (including outside storage tanks, etc.) to determine whether the above limits of Specification 3.25 have been exceeded.
- b. If the above limits have been exceeded, prepare and submit to the Commission within 30 days, a Special Report pursuant to Specification 6.9.5 that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405(c), shall include an analysis that estimates the radiation exposure (dose) to a HEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, in a calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations.
- c. If the estimated dose (s) exceed the above limits, and if the release condition resulting in the violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provision of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until
- staff action on the request is complete.
EMI 155- This specification is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the numerical guides for design objective doses of Appendix I or exceeds the reporting levels for the Radiological Environmental Monitoring Program. For the Rancho Seco site it is unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the plant remains within twice the numerical guides for design objectives of 10 CFR 50 Appendix I and if
~ direct radiation (outside storage tanks, etc.) is kept small. The Special Proposed Amendment No. 155, Revision 2, Supplement 1 3-90
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.25 (Continued)
BAsn (Continued) 155- Report will describe a course of action which should result in the limitation of the dose to a HEMBER OF THE PUBLIC for a calendar year to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the HEMBER 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 HEMBER OF THE PUBLIC during any period in which he/she is
- engaged in carrying out any operation which is part of the uraniun fuel cycle.
Proposed Amendment No. 155, Revision 2, Supplement 1 3-91
FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921, REV. 2, SUPP. 1 PROPOSED AMENDHENT NO. 155, REV. 2 PAGE 78
- 23. Existina Soecification: (Cont.)
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 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the HEMBER 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 Soecification:
3.25 FUEL CYCLE DOSE The dose or dose commitment to any real MEMBER OF THE PUBLIC due to releases of radioactive material in gaseous and liquid effluents and to direct radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem) in a calendar year.
Applicability At all times Alti al
- 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
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FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921, REV. 2, SUPP. 1 PR0 POSED AMENDHENT NO. 155, REV. 2 PAGE 79
- 23. New Soecification: (Cont.)
releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR part 20.405(c), shall include an analysis that estimates the radiation exposure (dose) to a HEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, in a calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations.
- 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 Honitoring Program. For the Rancho Seco site it is unlikely that the resultant dose to a HEHBER 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 HEMBER OF THE PUBLIC for a calendar year to within the 40 CFR 190 limits. For the purpose of the Special Report, it may be assumed that the dose commitment to the HEMBER 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 MEH3ER OF THE PUBLIC is evaluated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided l the release conditions resulting in violation of 40 CFR 190 have not l
already been corrected), in accordance with the provision
- of 40 CFR 190 l
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-3 MINIMUM SAMPLING FREQUENCY Item Check Frecuency
- 1. Reactor coolant a. Radio-chemical analysis (l) H 155-~ E determination (3)l4)(6) Semiannually
- b. Gross activity (1) (3) 3/ week
- c. Tritium radioactivity M
- d. Chemistry (C1 and 02 ) 3/ week
- e. Boron concentration 2/ week
- f. Fluoride H
- 2. Borated water Boron concentration (5) H and after each storage tank makeup water sample
- 3. Core flooding tank Boron concentration (3) H and after each water sample makeup
- 4. Spent fuel storage Boron concentration M and after each water sample makeup
- 5. Secondary coolant a. Gross activity (3) Weekly
- b. Iodine analysis (2)(3) Heekly 155-~ 6. Concentrated boric Boron concentration (5) 2/ week and after acid tank each makeup 155- 7. Spray additive NaOH concentration (3) Q and after tank each makeup
- 8. Cooling Tower Gross activity (3) M
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water Proposed Amendment No. 155, Revision 2, Supplement 1 4-9
155-RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-3 MINIMUM SAMPLING FREQUENCY Table Notation (l)Hhen radioactivity level is greater than 20 percent of the limi.ts of Technical Specification 3.1.4, the sampling frequency shall be increased to a minimum of once each day.
(2)Hhen gross activity increases by a factor of two above normal, an iodine analysis will be made and performed thereafter when the gross activity increases by ten percent.
(3)Not performed during cold shutdown.
(4)E determination will he started when a gross activity analysis indicates greater than 10pCi/gm. E will be redetermined each 10pCi/gm increase in gross activity. A radio chemical analysis for this purpose shall consist of a quantitative measurement of 95% of radionuclides in reactor coolant with half lives of >20 minutes.
(5)Not required during periods when systems are shutdown for maintenance.
(6) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since the reactor was last subtritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
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l Proposed Amendment No. 155, Revision 2, Supplement 1 1
4-9a
FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921, REV. 2, SUPP. 1 PROPOSED AMENDHENT NO. 155, REV. 2 PAGE 848 TABLE 4.1-3 HINIMUM SAMPLING FREQUENCY
- 25. New Soecification:
Item Check Freauency
- 1. Reactor coolant a. Radio-chemical analysis (I) H E determination (3)t4)(6) Semiannually
- b. Gross activity (1) (3) 3/ week
- c. Tritium radioactivity H
- d. Chemistry (C1 and 02 ) 3/ week
- e. Boron concentration 2/ week
- f. Fluoride H
- 2. Borated water Boron concentration (5) H and after each storage tank makeup .
water sample
- 3. Core flooding tank Boron concentration (3) H and after each water sample makeup
- 4. Spent fuel storage Boron concentration H and after each water sample makeup
- 5. Secondary coolant a. Gross activity (3) Heekly
- b. Iodine analysis (2)(3) Weekly
- 6. Concentrated boric Boron concentration (5) 2/ week and after acid tank each makeup i
l 7. Spray additive NaOH concentration (3) Q and after tank each makeup
- 8. Cooling Tower Gross activity (3) H water l
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FACILITY CHA"GE SAFETY ANALYSIS LOG NO. 921, REV. 2, SUPP. 1 PROPOSED AMENDHENT NO. 155, REV. 2 PAGE 84c
- 25. New Soecification: (Cont.)
TABLE 4.1-3 MINIMUM SAMPLING FREQUENCY Table Notation (I)Hhen radioactivity level is greater than 20 percent of the limits of Technical Specification 3.1.4, the sampling frequency shall be increased to a minimum of once eacn day.
(2)Hhen gross activity ir. creases by a factor of two above normal, an todine analysis will be made and performed thereafter when the gross activity increases by ten percent.
(3)Not performed during cold shutdown.
(4)5 determination will be started when a gross activity analysis indicates greater than 10pCi/gm. E will be redetermined each 10pCi/gm increase in gross activity. A radio chemical analysis for this purpose shall consist of a quantitative measurement of 95% of radionuclides in reactor coolant with half lives of >20 minutes.
(5)Not required during periods when systems are shutdown for maintenance.
(6) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since the reactor was last subtritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
Discussion:
Changes to Table 4.1-3 represent conformance to Standard Technical Specifications of NUREG-0800.
Footnotes (5) and (6) were added to accommodate system shutdown for maintenance and to ensure reactor stability for radio-chemical analysis and E determination at power, respectively.
Cooling tower blowdown is the same as cooling tower water. Eactivityis measured in pCi/gm not pci/ml.
Haste Gas Decay Tank, Auxiliary Building Plant Vent, and Purge Vent sampling frequencies are addressed in Table 4.22-1 and should not be in Table 4.1 .3 The existing specification footnote (4), regarding 10% of 10 CFR 20 limits, is deleted because it represents the old concentration release rate requirement. Specification 3.18.la is now the controlling requirement.
155-RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.20-1 (Continued)
TABLE NOTATION (1) The CHANNEL TEST shall also demonstrate that automatic termination of the purge and control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip retpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
(2) The CHANNEL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
(3) The INSTRUMENT CHANNEL CALIBRATION shall be performed using one or more reference standards.
(4) A check shall be performed prior to each release.
(5) A check shall be performed prior to each release via a Hasto Gas Decay Tank (s).
(6) To be performed when device is accessible and conditions do not pose a personnei safety hazard (i.e., potential main steam safety actuation).
(7) The CHANNEL TEST shall also demonstrate that the Haste Gas System automatically isolates and that control room annunciation occurs if any of the following conditions exist:
- 1. Instrument indictes measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
t l 4. Instrument controls not set in operate mode, i -
t Proposed Amendment No. 155, Revision 2, Supplement 1 4-68
FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921, REV. 2, SUPP. 1 PROPOSED AMENDHENT NO. 155, REV. 2 PAGE 97
- 29. New Specification: (Cont.)
TABLE 4.20-1 (Continued)
Table Notation (1) The CHANNEL TEST shall also demonstrate that automatic termination of the purge 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 do not set in operate mode.
(2) The CHANNEL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
(3) The INSTRUMENT CHANNEL CALIBRATION shall be performed using one or more reference standards.
(4) A check shall be performed prior to each release.
(5) A check shall be performed to each release via a Haste Gas Decay Tank (s).
(6) To be performed when device is accessible and conditions do not pose a personnel safety hazard (i.e., potential main steam safety actuation).
(7) The CHANNEL TEST shall also demonstrate that the Haste Gas System automatically isolates and that control room annunciation occurs if any of the following conditions exist:
- 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.
FACILITY CHANGE SAFETY ANALYSIS LOG N0. 921, REV. 2, SUPP. 1 PROPOSED AMENDMENT NO. 155, REV. 2 PAGE 97a
- 29. New Soecification: (Cont.)
TABLE 4.20-1 (Continued)
Table Notation (Continued)
Discussion:
The changes to Table 4.20-1 are made to be pursuant with the Standard Radiological Environmental Technical Specifications. Table Notation (6) is added to ensure safety of technicians performiong the channel test on the Auxiliary Building stack flow-rate device due to the device's location.
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, 155-RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.21-2 RADI0 ACTIVE LIQUID HASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Analysis Type of Activity Lower Limit Liquid Release Frequency Frequency Analysis (c) Of Detect'.on Type (LLD)
(uCi/ml)(a)
A.BatchHastegg- Each Batch Composite (d) Principal 4E-9 lease Tanks ( > P M Gamma Emitting Nuclides (c)
H-3 lE-5 Gross Alpha 1E-7 Sr-89, Sr-90 3E-8 Proposed Amendment No. 155, Revision 2, Supplement 1 4-72b
, - - - , - - , - . . < - - . , , - - , - , . ~ . . , - - . . . - _,n. ,
FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921, REV. 2, SUPP. 1 PROPOSED AMENDHENT NO. 155, REV. 2 SUPP. 1 PAGE 111
- 32. New Soecification (Cont.)
Table 4.21-2 RADI0 ACTIVE LIQUID HASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Analysis Type of Activity Lower Limit Liquid Release Frequency Frequency Analysis (c) Of Detection Type (LLD)
(uCi/ml)(a)
Each Batch Composite (d) Principal 4E-9 A. Batch Hasteb)(Re-lease Tanks P H Gamma Emitting Nuclides (c)
H-3 lE-5 Gross Alpha lE-7 Sr-89, Sr-90 3E-8 i
FACILITYCHANGESAFETYANALYSIS LOG N0. 921 REV. 2, SUPP. 1 PROPOSED AMENDMENT NO. 155, REV. 2, SUPP. 1 PAGE 142
- 43. Existing _SpAcification:
Table 4.26-1 (Continued)
MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) a, d Table Notation
- a. The LLD is defined in the ODCH.
Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLO'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.
- 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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e FACILITY CHANGE SAFETY ANALYSIS LOG NO. 921 REV. 2
- PROPOSED AMENDMENT NO.155 REVISION 2 PAGE 143 ,
- 43. New $pecification:
RANCHO SECO UNIT 1 TECHNIC #L SPECIFICATIONS .
Surveillance Standards Table 4.26-1 MAXIMUM VALUES FOR THE LOWEP. LIMITS OF DETECTION (LLD)a, d Water Airborne Particulate Fish Hilk Food Products Mud and Silt Analysis (pCi/1) or Gases (pCi/m )
3 (pCi/kg, wet) (pCi/1) (pCi/kg, wet) (pCi/kg, wet) gross beta 4(b) 1 x 10-2 3H 2000 (1000(b))
54Mn 15 130 59Fe 30 260 58Co 15 130 150 60Co 15 130 150 652n 30 260 95Zr-Nb 15 (e) 131I 1 (b) 7 x 10-2 1 60 134Cs 15 (10(b)) 1 x 10-2(c) 130 15 60 150 137Cs 18 (10(b)) 1 x 10-2(c) 150 18 80 180 I40Ba-La 15 te) 15 (e) _
Discussion:
Changes to Table 4.26-1 represent District confonr.ance to the Standard RETS. Lower maximum LLD values are defined for Cesium deter. tion in drinking water and milk. In addition, LLD values are established for the liquid effluent pathway in mud and silt fer 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 silt effluent pathway model did not take into account long-term buildup of concentrations of radionuclida, la bottom sediments, thereby imparting dose to ingesting aouatic foods (bottom-feeding fish).
The values for the LLD's come from a November 1979 NRC Branch Technical Position.
RANCHO SECO UNIT 1
. TECHNICAL SPECIFICATIONS LIST OF TABLES (Continued)
Table Past 4.14-1 DESIGNATED SAFETY RELATED HYDRAULIC SNUBBERS FUNCTIONALLY 4-47d e TESTED ONLY AS REQUIRED BY THE SNUBBER SEAL REPLACEMENT PROGRAM 4.17-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED 4-56 DURING INSERVICE INSPECTION a.17-2A STEAM GENERATOR TUBE INSPECTION 4-57 4.17-2B STEAM GENERATOR TUBE INSPECTIr.; (;eECint LIMITED AREA) 4-57a 4.17-3 OTSG AUXILIARY FEEDHATER HEATER SURVEILLANCE 4-57b,c 4.19-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADIOACTIVE GASEQUS EFFLUENT MONITORIHG INSTRUMENTATION 4-66 SURVEILLANCE REQUIREMENTS 4.21-1 RADI0 ACTIVE LIQUID RASTE SAMPLING AND ANALYSIS PROGRAM 4-70 155-~ 4.21-2 RADI0 ACTIVE LIQUID HASTE SAMPLING AND ANALYSIS PROGRAM 4-72b 4.22-1 RADI0 ACTIVE GASEQUS HASTE SAMPLING AND ANALYSIS PROGRAM 4-75 4.26-1 MAXIMUM VALUES FOR THE LOHER LIMITS OF DETECTION (LLD) 4-84 4.28-1 EXPLOSIVE GAS MIXTURE INSTRUMENTATION SURVEILLANCE 4-88 REQUIREMENTS 138-~ 6.2-1 MINIMUM SHIFT CREW COMPOSITION 6-2 Proposed Amendment No. 138, Rev. 2 Proposed Amendment No. 155 x
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