|
|---|
Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20155C6391998-10-26026 October 1998
Proposed Tech Specs Tables 4.1-1 & 4.1-2,revised Surveillance Frequency of Refueling Outage to 18 Months
ML20154C1021998-09-30030 September 1998
Proposed Tech Specs Changing UFSAR by Increasing Max Rod Internal Pressure in Spent Fuel Pool from 1,200 to 1,300 Psig
ML20236V9611998-07-27027 July 1998
Corrected TS Bases Pages 3.5-30b of 961211 & 3.16-2 of 970205,correcting Amend Numbers
ML20236T5171998-07-20020 July 1998
Proposed Tech Specs Re one-time Extension to Functional Test Frequencies for Hydraulic & Mechanical Snubbers
ML20236R1901998-07-16016 July 1998
Proposed Tech Specs Extending TS Surveillances TS 4.20.2(a), Table 4.20-1,items 1,2 & 4,TS 4.1-1,Table 4.1-1,items 8,9,10 & 11 & TS 4.5.1.2.1,item b(1)
ML16141B0771996-11-18018 November 1996
Errata to TS 3.7,which Removes Criteria for Battery & Battery Charger Specific Svc Tests
ML16141A9291995-06-29029 June 1995
Proposed Tech Specs Re Mgt Positions Authorized to Approve Such Items as Procedures & Procedure Changes,Station Mods, TS Amends & Reportable Events
ML16141A8111994-04-20020 April 1994
Proposed Tech Spec Table 4.1-2, Min Equipment Test Frequency
ML16141A7981994-04-12012 April 1994
Proposed Tech Specs Re Technical Review & Control Activities
ML15261A4281994-02-24024 February 1994
Proposed TS 4.6, Emergency Power Periodic Testing
ML20044C9581993-05-0404 May 1993
Proposed TS 4.7.1, Control Rod Trip Time Test
ML18032A3441987-05-29029 May 1987
Proposed Tech Specs,Clarifying Trip Level Setting in Table 3.2.A for Standby Gas Treatment Sys Relative Humidity Heater
ML15264A2511984-09-11011 September 1984
Proposed Tech Specs Supporting Operation of Facility at full-rated Power During Cycle 9
ML15223A8931983-05-19019 May 1983
Proposed Tech Spec Changes Re Core Protection Safety Limits, Protective Sys Max Allowable Setpoints & Rod Position Limits
ML16134A6761982-08-11011 August 1982
Proposed Tech Spec Revisions Re Reload Design Calculations for Cycle 7
ML16134A6731982-05-0303 May 1982
Proposed Tech Spec Changes Re Core Protection Safety Limits, Protective Sys Max Allowable Setpoints & Rod Position Limits
ML15223A7671982-01-12012 January 1982
Proposed Revisions to Tech Spec Section 3.1.2 Re Heatup Cooldown & Inservice Test Limitations for RCS
ML16148A4421981-11-13013 November 1981
Proposed Tech Spec Revision Re Core Protection Safety Limits Protective Sys Max Allowable Setpoints & Rod Position Limits
ML16148A4351981-10-28028 October 1981
Proposed Revision to Tech Spec Figure 3.5.2-4B2,allowing Cycle 5 to Run at 100% Full Power W/Axial Power Shaping Rods Fully Inserted
ML16148A4281981-08-19019 August 1981
Proposed Revision to Tech Spec Figures 3.5-16a,3.5.19a, 3.5-22 & 3.5-25a Re Extension of Operating Limits
ML15223A7321981-05-29029 May 1981
Proposed Tech Specs 2.1-2,2.1-3,2.1-7,2.3-5,3.2-1,3.2-2, 3.3-1,3.3-2,3.3-3,3.3-4,3.3-5,3.3-6,3.5-9,3.5-10,3.5-15, 3.5-15a & b,3.5-18,3.5-18a,b,c,d & e,3.5-21,3.5-21a & B, 3.5-24,3.5-24a & b,3.8-2 & 3.8-3 Re Core Protection
ML15112B0161981-04-20020 April 1981
Revised Tech Specs Pages Per Order Modifying License, Requiring Periodic Surveillance Over Life of Plant & Specifying Limiting Conditions for Operation of Primary Coolant Sys Pressure Isolation Valves
ML15112A9721980-10-24024 October 1980
Revised Tech Spec Pages Per 801024 Order for Mod of Licenses Re Environ Qualification of safety-related Electrical Equipment
ML16134A6651980-08-25025 August 1980
Proposed Tech Specs Revision for Cycle 6
ML16148A3351980-07-16016 July 1980
Proposed Revision to Tech Spec 3.3.1.c Allowing Continued Operation of Unit 2 of Full Rated Power While Maint Continues on HPI Pump Until 800718
ML19318B9731980-06-24024 June 1980
Proposed Tech Spec Interpreting Term Operable as Applied to Various Tech Spec Requirements
ML19317H3061980-04-10010 April 1980
Model Tech Specs for PWRs & BWRs
ML19317H3191980-02-21021 February 1980
Nuclear Data Link Spec,Revision 0,Draft 5
ML15238B2061980-01-15015 January 1980
Model Tech Specs for Fire Protection Program
ML16148A2691979-11-16016 November 1979
Proposed Changes to Tech Spec Pages 2.1,2.3,3.2 & 3.5. Changes Affect Core Protection Limits,Reactor Protective Sys Max Allowable Setpoints & Vol Requirements for Borated Water Storage Tank
ML19317D2701978-09-25025 September 1978
Proposed Tech Spec 3.5.2 Re Control Rod Group & Power Distribution Limits & Table 4.1-2 Re Min Equipment Test Frequency
ML19317D2621978-09-18018 September 1978
Proposed Revisions to Tech Specs 2.1,2.3,3.2 & 3.5 Re Core Protection Safety Limits & Protective Sys Max Allowable Setpoints
ML19317D2731978-09-0606 September 1978
Revised Tech Spec Page,Figure 2.3-2A,re Protective Sys Max Allowable Setpoints
ML19317D2491978-08-22022 August 1978
Proposed Revision to Tech Specs 4.18 Re Hydraulic Shock Suppressors (Snubbers)
ML19322B9961978-08-21021 August 1978
Proposed Revision to Tech Spec 2.3 Re Cycle 5
ML19329A2791978-08-0707 August 1978
Facility Tech Specs 3.9.9 Through 3.9.11 to Control Waste Water Pond Radioactivity
ML19312B7931978-07-17017 July 1978
Proposed Tech Spec 3.1.6.4,changing Steam Generator Leak Rate Limit
ML19308D6271978-06-28028 June 1978
Tech Spec Change Request Re Paragraph 2.B(6),stipulating That Byproduct & SNM Associated W/Four Fuel Assemblies Acquired by Fl Power Corp from Duke Power Co Previously Irradiated in Oconee 1 May Be Possessed
ML19317D2301978-06-26026 June 1978
Proposed Tech Specs 2.1,2.3,3.2,3.5 & 4.1 Required to Support Operation of Unit 1 at Full Rated Power During Cycle 5,including Core Protection Safety Limits & Protective Sys Mac Allowable Setpoint
ML19316A6501978-06-22022 June 1978
Proposed Replacement Page for Tech Spec 4.1-2 Re Min Equipment Test Frequency
ML19316A5381978-06-14014 June 1978
Proposed Changes to Tech Specs Re thermal-hydraulics Analysis.Revision to BAW-1486, Unit 3,Cycle 4 Reload Rept
ML19312B8161978-06-12012 June 1978
Proposed Tech Specs 3.8,4.4 & 4.6 Re Fuel Loading & Refueling,Structural Integrity & Emergency Power Periodic Testing
ML19317D2341978-06-0909 June 1978
Proposed Tech Spec 3.9 Deleting Requirements Not Applicable to Liquid Effluent Monitoring Sys Due to Installation of Offline Monitor
ML19317D2221978-06-0808 June 1978
Proposed Tech Spec 3.1 Allowing Max 1 Gallon Per Minute Leakage Through Steam Generator Tubes Prior to Initiation of Unit Shutdown
ML19317D2121978-06-0202 June 1978
Proposed Tech Spec 4.2 Allowing re-insp of Reactor Coolant Outlet Nozzles at Future Refueling Outage
ML19316A5271978-05-30030 May 1978
Proposed Revisions to Tech Specs 2.3,3.2 & 3.5.2.4 to Support Cycle 4 Operation at Full Power
ML19312B7971978-04-27027 April 1978
Proposed Tech Spec 6.4,incorporating Operating Procedure Requirements Re B&W Small Break ECCS Analysis
ML19312B8091978-04-20020 April 1978
Proposed Tech Spec 3.3 Incorporating New Tech Spec 3.3.8 Requiring Operability of Three HPI Pumps for Each Unit During Power Operation Above 60% Full Power
ML19317D2001978-03-20020 March 1978
Proposed Tech Spec 3.5 Including 6.03% Quadrant Power Tilt Limit & Provision for Notifying NRC If Tilt Exceeds 3.5%
ML19317D2131978-02-21021 February 1978
Proposed Tech Spec 3.1. Incorporating Revision to Pressurization,Heatup & Cooldown Limitations.B&W to Util Re Corrections to Errors Discovered in B&W Rept BAW-1436 Encl
1998-09-30
[Table view]
Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20211A9881999-08-18018 August 1999
Rev 5 to DPC Nuclear Security Training & Qualification Plan
ML20204B4141999-03-27027 March 1999
Revised Oconee Nuclear Station Selected Licensee Commitments, List of Effective Pages
ML20155C6391998-10-26026 October 1998
Proposed Tech Specs Tables 4.1-1 & 4.1-2,revised Surveillance Frequency of Refueling Outage to 18 Months
ML20154L7191998-10-0505 October 1998
Rev 8 to DPC Nuclear Security & Contingency Plan, for Oconee,Mcguire & Catawba Nuclear Stations
ML20154C1021998-09-30030 September 1998
Proposed Tech Specs Changing UFSAR by Increasing Max Rod Internal Pressure in Spent Fuel Pool from 1,200 to 1,300 Psig
ML20236V9611998-07-27027 July 1998
Corrected TS Bases Pages 3.5-30b of 961211 & 3.16-2 of 970205,correcting Amend Numbers
ML20236T5171998-07-20020 July 1998
Proposed Tech Specs Re one-time Extension to Functional Test Frequencies for Hydraulic & Mechanical Snubbers
ML20236R1901998-07-16016 July 1998
Proposed Tech Specs Extending TS Surveillances TS 4.20.2(a), Table 4.20-1,items 1,2 & 4,TS 4.1-1,Table 4.1-1,items 8,9,10 & 11 & TS 4.5.1.2.1,item b(1)
ML20217F2841998-04-20020 April 1998
Rev 7 to DPC Nuclear Security & Contingency Plan, for Oconee,Mcguire & Catawba Nuclear Stations
ML20141F1521997-06-25025 June 1997
Rev 4 to Nuclear Security Training & Qualification Plan
ML16141B0771996-11-18018 November 1996
Errata to TS 3.7,which Removes Criteria for Battery & Battery Charger Specific Svc Tests
ML20134K5401996-10-31031 October 1996
Rev 6 to Chemistry Manual 5.1, Emergency Response Guidelines
ML20117J0181996-08-15015 August 1996
Revised Chapter 16 of Oconee Selected Licensee Commitments Manual
ML20095H1291995-12-0505 December 1995
Rev to ONS Selected Licensee Commitments (SLC) Manual, Revising SLC 16.6.1, Containment Leakage Tests to Reflect Current Plant Configuration & Update Testing Info
ML20091P2461995-08-21021 August 1995
Rev to ONS Selected Licensee Commitments Manual
ML16141A9291995-06-29029 June 1995
Proposed Tech Specs Re Mgt Positions Authorized to Approve Such Items as Procedures & Procedure Changes,Station Mods, TS Amends & Reportable Events
ML16141A8581995-01-0909 January 1995
Revs to Oconee Selected Licensee Commitments Manual
ML16141A8111994-04-20020 April 1994
Proposed Tech Spec Table 4.1-2, Min Equipment Test Frequency
ML16141A7981994-04-12012 April 1994
Proposed Tech Specs Re Technical Review & Control Activities
ML15261A4281994-02-24024 February 1994
Proposed TS 4.6, Emergency Power Periodic Testing
ML15224A3521993-10-26026 October 1993
Safety Assurance Directive 6.1, Oconee Nuclear Site Safety Assurance Emergency Response Organization
ML20044C9581993-05-0404 May 1993
Proposed TS 4.7.1, Control Rod Trip Time Test
ML20045F0721993-04-0707 April 1993
Rev 9 to Corporate Process Control Program Manual
ML20097D9451992-05-28028 May 1992
Rev 11 to Training & Qualification Plan
ML20096C2561992-04-30030 April 1992
Public Version of Revised Crisis Mgt Implementing Procedures (Cmip),Including Rev 46 to CMIP-1,Rev 39 to CMIP-4,Rev 45 to CMIP-5,Rev 49 to CMIP-6,Rev 48 to CMIP-7,Rev 42 to CMIP-9 & Rev 3 to CMIP-15
ML20096D5451992-04-0707 April 1992
Public Version of Revised Crisis Mgt Implementing Procedures (Cmip),Including Rev 45 to CMIP-1,Rev 27 to CMIP-13 & Notification of Deletion of CMIP-8.Procedure CMIP-8 Reserved for Future Use
ML16131A5241992-03-0101 March 1992
Rev 35 to ODCM Generic Section
ML20092M5891992-02-0606 February 1992
Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 43 to CMIP-1,Rev 43 to CMIP-5,Rev 33 to CMIP-8,Rev 25 to CMIP-13,Rev 6 to CMIP-18,Rev 4 to CMIP-22 & Rev 38 to CMIP-4
ML20094H2391992-01-0101 January 1992
Rev 33 to McGuire Nuclear Station Odcm
ML20094H2511992-01-0101 January 1992
Rev 34 to Catawba Nuclear Station Odcm
ML16131A5221992-01-0101 January 1992
Rev 32 to Oconee Nuclear Station Odcm
ML20087F3931991-12-11011 December 1991
Public Version of Rev 13 to Crisis Mgt Implementing Procedure CMIP-11, Emergency Classification - Mcguire. W/ 920121 Release Memo
ML20086K8831991-11-18018 November 1991
Public Version of Revised Crisis Mgt Implementing Procedures (Cmips),Including Rev 42 to CMIP-1,Rev 29 to CMIP-2,Rev 42 to CMIP-5,Rev 12 to CMIP-11 & Rev 37 to CMIP-21
ML20086G7711991-10-16016 October 1991
Public Version of Revs to Crisis Mgt Implementing Procedures (Cmip),Including Rev 41 to CMIP-1,rev 37 to CMIP-4,rev 41 to CMIP-5,rev 46 to CMIP-6 & CMIP-7,rev 32 to CMIP-8,rev 40 to CMIP-9,delete CMIP-12 & Rev 24 to CMIP-13
ML20082C7441991-06-11011 June 1991
Public Version of Rev 13 to Crisis Mgt Implementing Procedure CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20081J9841991-06-10010 June 1991
Rev 3 to EDA-1, Procedure for Estimating Food Chain Doses Under Post-Accident Conditions
ML20076A7241991-06-10010 June 1991
Public Version of Crisis Mgt Implementing Procedures, Including Rev 39 to CMIP-1,rev 28a to CMIP-2,rev 44 to CMIP-7,rev 39 to CMIP-9 & Rev 10 to CMIP-11
ML20081K0101991-06-0606 June 1991
Rev 8 to EDA-3, Offsite Dose Projections for McGuire Nuclear Station
ML20072S0521991-03-15015 March 1991
Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 9 to CMIP-11, Classification of Emergency for McGuire Nuclear Station & Rev 11 to CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML16148A9621991-02-13013 February 1991
Public Version of Rev 10 to CMIP-12, Crisis Mgt Implementing Procedure
ML20066G3621991-02-0101 February 1991
Public Version of Crisis Mgt Implementing Procedures, Including Rev 28 to CMIP-2,Rev 34 to CMIP-4,Rev 38 to CMIP-5,Rev 43 to CMIP-6,Rev 42 to CMIP-7,Rev 29 to CMIP-8, Rev 37 to CMIP-9 & Rev 34 to CMIP-21
ML20082P7611991-01-0101 January 1991
Rev 30 to Odcm,Catawba Nuclear Station
ML15217A1291991-01-0101 January 1991
Rev 29 to Odcm,Oconee Nuclear Station
ML20072S9621991-01-0101 January 1991
Public Version of Rev 12 to Crisis Mgt Implementing Procedure CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20082P7711991-01-0101 January 1991
Rev 31 to Odcm,Mcguire Nuclear Station
ML20072P9621990-11-0808 November 1990
Rev 9 to Crisis Mgt Implementing Procedure CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20028H2211990-10-31031 October 1990
Public Versions of Revised Crisis Mgt Implementing Procedures,Including Rev 37 to CMIP-1,Rev 27a to CMIP-2,Rev 33 to CMIP-4,Rev 37 to CMIP-5 & Rev 42 to CMIP-6
ML20059F4301990-08-22022 August 1990
Public Version of Rev 27 to Crisis Mgt Implementing Procedure CMIP-2, News Group Plan
ML20063Q2721990-08-14014 August 1990
Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 36 to CMIP-1,Rev 32 to CMIP-4,Rev 36 to CMIP-5,Rev 41 to CMIP-6,Rev 40 to CMIP-7,Rev 27 to CMIP-8,Rev 35 to CMIP-9 & Rev 7 to CMIP-11
ML20043F4841990-05-23023 May 1990
Public Version of Crisis Mgt Implementing Procedures, Consisting of Rev 20 to CMIP-13 & Deletion of CMIP-15 & CMIP-17
1999-08-18
[Table view] |
Text
.
s
.q
, 1.*8 DOSE EQUIVALENT - I-131 The Dose Equivalent I-131 shall be that concentration of I-131 (uci/ gram) which alone would produce the same thyroid dose as the cuantity 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".
1.9 5 - AVERAGE DISINTEGRATION ENERGY the 5 Average Disintegration Energy shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and ga==a energies per disintegration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
i i
I 191125007b l-5 i
3.1. 4,
Reactor Coolant Svstem Activity Soecification 3.1.4.1 The specific activity of the reactor coolant system shall not exceed 3.5 pCi/ gram dose equivalent I-131 except as provided in Specification 3.1.4.2.
3.1.4.2 If the specific activity of the reactor coolant system is greater than 3.5 uCi/ gram but less than 60 uCi/ gram dose equivalent I-131, power operation =ay continue for periods of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
3.1.4.3 The specific activity of the ?.eactor Coolant System shall not exceed 311/E uC1/ gram.
3.1.4.4 If the conditions of Specifications 3.1.4.2 or 3.1.4.3 are not met, the reactor shallbe in a hot shutdown condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
A reportable occurrence shall be submitted to the Commission pursuant to Specification 6.6.2.1.b and shall contain the following information:
a.
Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded.
b.
Fuel burnup by core region
~
Cleanup flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first c.
sample in which the limit was exceeded d.
History of degassing operations, if any, starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> i
prior to the first sample in which the limit was exceeded The time duration when the specific activity of the reactor _
e.
coolant exceeded 3.3 uCi/gres dose equivalent I-131 or 311/E uCi/ gram.
Bases The limitations on the specific activity of the reactor coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriate fraction of the Part 100 limits following a steam generator tube rupture accident. The steam generator tube rupture was~ analyzed as a separate accident and in conjunction with a steam line break accident.
The analyses considered the effects of the " iodine spike" associated with the accident transient and also considered the compounding effects of a preexisting iodine spike caused by seme prior transient.
The following lodine spiking model, empirically developed from operating data, was used to calculate the curies of iodine entering the secondary system after the tube rupture:
3.1-4
~ ( i+1 i} + C0(f )(fo) e~
i+1(1-e- ( i+1~ i)
~
~
-C
=C e
g41 where:
C
= I-131 activity in the reactor coolant at time e, uC1/ gram 1
g time after start of accident, min.
t
=
g L, = total removal rate of-I-131, min ~
R
= release rate of I-131 into the reactor coolant, Ci/ min.
A
= radioactive decay constant for I-131, min ~
C*
= steady state I-131 concentration prior to transient =3.5 uCi/ gram L*
steady state removal rate of I-131, prior to transient
=
= 6.85 x 10-4 min-1 R*
steady state release rate of I-131 into the reactor coolant,
=
C1/ min Time After Transien't Hours Spiking Factor (ti)
(R/R*)
0-1 127 1-2 47 2-3 16 3-4 6
4-5 2.3 For the case of a preexisting iodine spike, the curies of iodine entering the secondary system were based on Ci at the time of the tube rupture (ti=0) being 60 uCi/ gram dose equivalent I-131.
An accident must occur in a very small " time window" following a power transient for the iodine concentration to be at 60 uC1/ gram since this high concentration exists for only a rela-tively short period of time following a transient.
The primary to secondary e
leak rate associated with a double-ended steam generator tube rupture accident was conservatively assumed to be a constant 435 gpm (43.6 lbs/sec).
All of the noble gas activity and 10% of the iodine activity in the leakage entering the secondary system is assumed to be present in the steam = ass release to the environment. Also assumed to be released to the environment is the iodine activity in 173,300 lbs of secondary coolant containing 0.1 uCi/ gram of dose equivalent I-131 (the maximum value per Tech Spec 3.13) and the iodine and noble gas activity-associated with a primary to secondary leak rate of 1 g;m (the maximum leak rate permitted per Tech Spec 3.1.6)'.
3.1-5
~
The site boundary doses were based on the zero to two-hour dispersion (X/Q), of 1.16 x 10-E (ec/m3 (per Section 2.3.2 of the FSAR).
factor at the site-boundarv 1609 m) corresponding to a ground release, s
The i.e.,
dose calculations are_ consistent with TID-14844, except for the conserva-tive assumption that E used to calculate the whole body dose includes both the beta and ga=ma energy whereas TID-14844, Reg. Guides 1.4,1.24, 1.25, and 1.77 only consider the ga=ma energy in calculating the whole body dose.
The resulting doses are:
2 Hour Site Boundarv Doses (Rem)
Thyroid Whole Body i
Steam Generator Tube Rupture with 2.8 0.34 Iodine Spike Steam Generator Tube Rupture with 14.0 0.86 Steam Line 3reak and Iodine Spike Steam Generator Tube Rupture with 16.1 0.34 Preexisting Iodine Spike i
Steam Generator Tube Rupture with 46.7 0.86 Steam Line 3reak and Preexisting Iodine Spike Power operation for time periods with the reactor coolant's specific activity > 3.3 uCi/ gram dose equivalent I-131, but less than 60 uCi/ gram dose equivalent I-131, acccmodates possible iodine spiking phenomenon whieb may occur following changes in thermal power.
Operation with specific activity levels exceeding 3.5 uCi/ gram dose equivalent I-131 but within the 60 uCi/ gram dose equivalent I-131 limit is restricted to periods not to exceed 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> since these activity levels increase the l
2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of 3 to 6 following a postulated steam generator tube rupture.
Reducing Tavg to < 5300F prevents the release of activity in the event of a steam generator tube rupture since the saturation pressure of the reactot coolant is below the lift pressure of the atmospheric steam relief valves.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
The information reported relative to iodine spiking will help to assess the parameters associated
.tth spiking phenomena. A reduction in frequency of isotopic analyses following power changes =ay be permissible when justified by. the data obtained.
3.1-6
3.13 SECONDARY SYSTCt ACTIVITY 1
Applicability Applies 'to the limiting conditions of secondary system activity for operation of the reactor.
Objective To limit the maximum secondary system activity.
Specification 3.13.1 The specific activity of.the secondary coolant system shall not exceed 0.10 uct/ gram dose equivalent I-131.
3.13.2 If the secondary coolant system specific activity exceeds 0.10 uCi/ gram dose equivalent I-131, the reactor shall be placed in a hot shutdown condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in a cold shutdown condition within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Bases The limitations on secondary system specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10CFR100 limits in the eyent of a steam line rupture.
This dose includes the effhets of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line.
These values are consistent with the assumptions used in the safety analyses.
a 1
t I
j 1
3.13-1 n
n a
,+m,-
y
l TAlli.E 4.1-3
\\
MINIMtJM SAMPLING FREQtIENCY Item Check Frequency
<l-1.
Reactor Coolant a.
Canuna Isotopic Analysis a.
Monthly
Itadlochemical Analysis for Sr 89, 90 h.
Monthly
Tritium c.
Monthly
Cross lieta or Cross Canuna Activity (1) d.
S times / week
- Chemistry (C1, F and,02) e.
S times / week
f.
lloron Concentration f.
2 times / week **
g.
Cross Alpha Activity g.
Monthly
E Determination (2) h.
Semi-annually *
luotopic Analysis for I-131 1.
5 times / week
Isotopic Analysis for Dose E<1uivalent J.
Once per 14 days
Isotopic Analysis for Iodine Including k.
See note (3)*
I-131, I-133 and 1-135 2.
llorated Water Storage Iloron Concentration Weekly
Tank Water Sample IWo 3.
Core Flooding Tank lloron Concentra t ion Monthly
Spent Fuel Pool Water lloron Concentration Monthly *** and af ter each makeup 5.
Secondary Coolant a.
Isotopic Analysis for Dose Equivalent a.
Weekly
Concentrated Iloric Acid lloron Concentration Twice Weekly
- Not Applicable.if reactor is in a cold si.utdown condition for a period exceeding the sampling frequency.
- Applicable only when fuel is.in the reactor.
- Applicably only when fuel is in wet storage in the spent fuel pool.
o
i TAlli.E 4.1-3 ConL.
ft_I N I ff U 11 S A H P I. I N C F it E Q U E N C Y Sensitivit y of Waste i
item Check Freilnency Analysis In I. ale
-7
-7.
1.uw Activity Waste
-a.
Canuna Isotopic Analysis
- a. Prio.r to release
- a. Canuna tincliiles <5x10 I*
5 Tank, Conilensate incluit ing Dissolveil of each l>atcli Dissolveil Cases <10 pC1/mi Test Tank, Hols le Cases Conilensate
_g Honitoring Tank, 13. Rattlochemical Analysis li. Monthly
- b. <10 pCi/mi
.s 1.rnnul ry-Ilo t Sr 89,90 Shower Tank
< 10 ~5 pCl/mi
- c. Tritium
- c. tionthly c.
al. Cross Al pha Ac t-I vi t-y al. Monthly 31. <10' pCl/ml
<10_4 pCi/cc (gases)
- 8. Waste Gas Decay a.
Camma luotopic Analysis a.
Prior to release a.
gg Tank.
of each liatch
<10 pC1/cc (particulates anti l otlines)
- h. Tritium
- h. Prior to release b.
<10~
pct /cc of each batch
- 9. Unit Vent Sampling a.
Io<line Spectrum
- a. Weekly a.
<10~
pCl/cc l
- h. Particulates 1
- 1) Canana luotopic Analysis
- 1) Weekly Composite
- 1) <10 pC1/cc
- 2) Cross Alpha Activity
- 2) Quarterly on a
- 2) <10
pCi/cc sample of one week luration
- 3) Raillochemical Analysis
- 3) Quarterly Composite
- 3) <10 pCl/cc Sr 119,90 h
i TAltl.E 4.1-3 Cont.
} -
H IN IHUM S A H p 1. I f3 C F lt E Q II E f3 C Y i
Sensitivity of Waste Item Check Frequency Analysis in I.ab
- c. Cases by Canuua Isotopic
- c. Weekly
- c. <10 ' pCl/cc Analysis
- 10. Keowee flydro Dam Heasure f.eakage Flod Rate Annually
. Dilution Flow I1.-Condenser Alr Ejector Heasure lodine parLitinn One time if and when primary partition Fac t o r Fac t o r in Condenser to seconlary leaks develop
)
-4
'12. Reactor lluilding
- a. Ga nuna Isot.opic Analysis a.
Each purge a.
<10 pci/cc (gases)
-10
<10 pCi/cc (particulates and fodines)
- b. TritInm
- h. Each I' urge b.
<10-pcl/cc
.s (1)
When radioact ivit y level is e,reater than 10 percent of the limits of Specification 3.1.4, the sampling frequency shall'he increased to a minimum of once each day.
(2)
E determination will be started when gross beta or gross ganuna activity analysis indicates greater,than 10 pC1/ml and will be redetermined for each 10 pCf /ml increase in gross beta or gross gamma activity analysis thereafter.
(3) The isotopic analysis of the reactor coolant for iodine including T-131, I 133, 1-134 and I-135 shall be performed once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> whenever the specific aelvity exceeds 3.5 pCi/ gram dose espilvalent I-131 or 311/E pCi/ gram.
One sample shall be analyzed between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a thermal power change exceeding 15 percent thermal
.)
power in a one hour period.
l h
n a
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