Proposed Changes to Tech Specs 3.1.2 Re Pressurization Heatup & Cooldown Limitations for Unit 3 Per B&W Jul 1977 rept,BAW-1438, Analysis of Capsule OCIII-A from Duke Power Co Oconee Unit 3 Reactor Vessel Matls Surveillance Program.| ML19317D211 |
| Person / Time |
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| Site: |
Oconee  |
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| Issue date: |
09/14/1977 |
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| From: |
DUKE POWER CO. |
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| To: |
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| Shared Package |
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| ML19317D210 |
List: |
|---|
| References |
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| NUDOCS 7911190602 |
| Download: ML19317D211 (9) |
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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)
ML16141A8111994-04-20020 April 1994
Proposed Tech Spec Table 4.1-2, Min Equipment Test Frequency.
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
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
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
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
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 780125 Ltr to Util Re Corrections to Errors Discovered in B&W Rept BAW-1436 Encl
ML19317D2211978-02-16016 February 1978
Proposed Tech Spec 2.3 Deleting Loss of One Pump Trip Setpoint,Outdated Info & Setpoints Associated W/Single Loop Operation
ML19340A2641978-02-0808 February 1978
Tech Specs 3.3.2 Through 3.3.5 for ECCS
ML19316A4801978-02-0101 February 1978
Proposed Changes to Tech Spec 3.7 Re Limiting Conditions for Operation & Surveillance Requirements for 125 Volt Distribution Sys
ML19317D2271978-01-23023 January 1978
Proposed Tech Spec 3.5 Incorporating Control Rod Position & Axial Imbalance Limits to Period After 100 Plus or Minus 10 Effective Full Power Days
ML19312B7891978-01-0303 January 1978
Proposed Tech Spec 2.3-9 & 10 Re Computer Software Used to Process Incore Detector Signal.Includes Modified power-imbalance Trip Setpoints to Account for Bias in Positive Imbalance Measured by Incore Detector Sys
ML19316A5081977-12-0202 December 1977
Amend to Tech Specs 3.9 Re Radwaste Discharge
ML19312B8001977-12-0202 December 1977
Proposed Tech Spec 6.6.2.1 Providing Requirement for Prompt Written Notification of Certain ROs by Telephone,Mailgram or Facsimile Transmission
ML19312B7771977-11-0909 November 1977
Proposed Tech Spec 3.5.2 Permitting Operation of Unit 1 During Cycle 4 in Unrodded Mode
ML19312B8061977-10-31031 October 1977
Proposed Tech Specs 6.6-1,-2,-3 & -4 Deleting Redundant Info Currently Being Reported in Annual Operating Rept
ML19312B8101977-10-26026 October 1977
Proposed Tech Specs 3.5-9 & 3.5-24a,deleting Existing Reactor Core Quadrant Power Tilt & Control Rod Position Limits & Instituting More Conservative Limits
ML19312B8151977-10-0707 October 1977
Proposed Changes to Tech Specs 3.7 & 4.6 Re Auxiliary Electrical Systems & Emergency Power Periodic Testing
ML19329A4011977-10-0606 October 1977
Revised Tech Specs,Table 4.1-3 Re Min Sampling Frequency
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
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
ML16141A8111994-04-20020 April 1994
Proposed Tech Spec Table 4.1-2, Min Equipment Test Frequency.
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
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
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
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
ML20081J9841991-06-10010 June 1991
Rev 3 to EDA-1, Procedure for Estimating Food Chain Doses Under Post-Accident Conditions
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
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
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
ML20043B2651990-05-0909 May 1990
Revised Crisis Mgt Implementing Procedures,Including Rev 35 to CMIP-1,Rev 26 to CMIP-2,Rev 31 to CMIP-4,Rev 35 to CMIP-5,Rev 40 to CMIP-6,Rev 39 to CMIP-7,Rev 26 to CMIP-8, Rev 33 to CMIP-9,Rev 2 to CMIP-14 & Rev 10 to CMIP-16
ML20043F4621990-04-20020 April 1990
Rev 5 to Oconee-specific Process Control manual.W/900606 Ltr
ML20006C0571990-01-18018 January 1990
Public Version of Rev 33 to Crisis Mgt Implementing Procedure CMIP-1, Recovery Manager & Immediate Staff & Rev 24 to CMIP-2, News Group Plan.
ML16152A8951990-01-0202 January 1990
Rev 33 to Public Version of Crisis Mgt Plan for Nuclear Stations.
ML15264A1571990-01-0202 January 1990
Revised Crisis Mgt Implementing Procedures,Including Rev 32 to CMIP-1,Rev 29 to CMIP-4,Rev 33 to CMIP-5,Rev 38 to CMIP-6,Rev 37 to CMIP-7,Rev 32 to CMIP-9,Rev 1 to CMIP-14 & Rev 30 to CMIP-21
ML20012A3801990-01-0101 January 1990
Rev 28 to, Offsite Dose Calculation Manual,Oconee,Mcguire & Catawba Nuclear Stations.
ML20012A3791990-01-0101 January 1990
Rev 27 to, Offsite Dose Calculation Manual,Oconee Nuclear Station.
ML20011D2441989-12-0101 December 1989
Crisis Mgt Implementing Procedures.
ML20012A3731989-11-15015 November 1989
Rev 4 to, Process Control Program Oconee Nuclear Station.
1999-08-18
[Table view] |
Text
O-O o 3.1:2 P_r_Erurizntien, Hutup, end Cooldown Limitetion, Specification 3.1.2.1 The reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited as follows:
Heatup:
Heatup rates and allowable coe.binations of pressure ari tempera-tures shall be limited in accordance with Figure 3.1.2-1A Unit 1 3.1.2-1B Unit 2 3.1.2-1C Unit 3. l Cooldown:
Ccoldown rates and allowable combinations of pressure and tempera-ture shall ?;e limited in accordance with Figure 3.1.2-2A Unit 1 3.1.2-2B Unit 2 3.1.2- 2C Unit 3. l 3.1.2.2 Leak Tests Leak test required by Specification 4.3 shall be conducted under the provisions of 3.1.2.1.
3.1.2.3 Hydro Tests For thermal steady state system hydro test the system may be pressurized to the limits set fo.rh in Specification 2.2 when there are fuel assemblies in the core under the provisions of 3.1.2; 1 and to ASME Code Section III limits when no fuel assem-blies are present provided the reactor coolant system is to the ri;;ht of and below the limit line in Figure 3.1.2- 3A Unit 1 3.1.2-3B Unit 2 3.1.2-3C Unit 3. l l
! 3.1.2.4 The secondary aide of the steam generator shall not be pressurizad above 237 psig if the temperature of the vessel shell is below 110 F.
3.1.2.5 The pressurizer heatup and cooldown rates shall not exceed 1000F/hr.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 4100F.
3.1.2.6 Pressurization heatup and cooldown limitations and hydro test limits shall be updated based on the results of the reactor vessel materials surveillance program described in Specification 4.2.9. These revised limits shall be submitted to the NRC at least 90 days prior to ex-ceeding four (Unit 1) effective full power years of operation.
six (Unit 2) eight (Unit 3) l
~'
19111goQOQ
Bares - Unita'1, 2 and 3 All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, startup and s.hutdown operations, and inservice leak and hydrostatic tests.
The various categories of load cycles used for design purposes are provided in table 4.8 of the FSAR.
The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10CFR50. Results of this analysis, including the actual pressure-temperature limitations of the reactor coolant pressure boundary, are given in BAW-1421(1), BAW-1437(2) and BAW-1438(3). l The figures specified in 3.1.2.1, 3.1.2.2 and 3.1.2.3 present the pressure-temperature limit curves for normal heatup, normal cooldown and hydrostatic test, respectively. The limit curves are anplicable up to the indicated effective full power years of operation. These curves are adjusted by 25 psi and 100F for possible errors in the pressure and temperature sensing instru-ments. The pressure limit is also adjusted for the pressure differential between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations.
The pressure-temperature limit lines shown on the figure specified in 3.1.2.1 for reactor criticality and on the figure specified in 3.1.2.3 for hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inser-vice hydrostatic *.esting.
The actual shift in RTypT of the beltline region material will be established periodically durir.g operation by removing the evaluating, in accordance with Appendix H to 10CFR50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core region.
The limitation on steam generator pressure and temperature provide protection against nonductile failure of the secondary side of the steam generator. At metal temperatures lower than the RTNDT of +600F, the protection against non-ductile failure if achieved by limiting the secondary coolant pressure to 20 percent of the preoperational system hydrostatic test pressure. The I limitations of 1100F and 237 psig are based on the highest estimated RTNDT of +400F and the preoperational system hydrostatic test pressure of 1312 psig.
The average metal temperature is assunad to be equal to or greater than the coolant temperature. The limitations include margins of 25 psi and 100F for possible instrument error.
The spray temperature difference is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.
l 3.1-3a
')
];
REFERENCES (1) Analysis of Capsule OCl-F from Duke Power Company Oconee Unit 1 Reactor Vessel Materials Surveillance Program, BAW-1421 Rev. 1, September 1975.
(2) Analysis of Capsule OC2-lC from Duke Power Company Oconee Unit 2 Reactor Vessel Materials Surveillance Program, BAW-1437, April, 1977.
(3) Analysis of Capsule OCIII-A from Duke Power Company Oconee Unit 3 Reactor Vessel Materials Surveillance Program, BAW-1438, July, 1977.
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a E 2254 321 283 656 g i 323
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' 3 O setessa las palet el systes pressure esasuse:aat aat the pressure en Ins
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- ele lastrument asses (25 pse aus 10*f).
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- g E i n & scanlastness are to the sagat el tne oi. ..celity limit.
Q 3 10P6 Applicaele let Neetup Crill6ality Llall
[ {q) E 2 Bates up to 100f/a g3 g, ggg -
(< 68'l la any l/2 neut I
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, UNIT 3 REACTOR COOLANT SYSTEM NORMAL OPERATION llEATUP LIMITATIONS
- APPLICABLE FOR FIRST 8.0 EFPY 4
1 OCONEE NUCLEAR STATION
3468 y
- 1. Bass las Becas usat Baseval (SH) $sstes as aparatlag esta se SC rusos saaretsag.
2284 - tae sedicates 888 system retute toeperature to the seastst vessel small as used.
- 2. A mastaus step temperature canage of it'f as allesania ines sessenes all SC rumps f ree spesatsen ella las Oli $gstes operatlag. Ice ste:I 'espesatuss caange is estaaed 2048
- as ese SC isap. (Frees te stepping all at fusps) an.ius sons au setura insp. (attes stepping all SC pumps). The 188'f/h8 tamp secrease as ellesaale tota talere and af ter tee step taap caenge.
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- . seone; applicat.le coolde sates -
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- l gQOJ 408 and to the altut of the liest .ustets). Ine limit curves k_d include tue pressure clifers tssi motseem the pelat si systes pfessuse seasurtecat anJ tas pesssuse en the reacter gc 288 - vessel regles centselling time liset curve. laty lacla4s as ot esit ca. it n peo) a..iussa .a in ei suoi,le. ,essiois insa -sai s.,e,.
, , s
> I 288 328 368 488 128 188 284 244 48 84 C7 ladicates Osac';. Caetatt leaperature. *f EP) um (7 --
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P' d REACTOR COOLANT SYSTEM NORMAL OPERATION C001.00WN LIMITATIONS
- APPLICABLE FOR FIRST 8.0 EFPY OCONEE NUCLEAR STATION Figure 3.1.2-2C
2000 Pssssuse. .'cey ,
I E Fatet pse F
lett
- A lla St 8 550 132 C 558 255 Apsticante ter nestup rates et 2380 -
B 1433 264 < ggg.yjn, y E 2500 321
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- . ,s .e.)
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@ 484 Ine acceptante piastuse temposature coatisations are teles and to the sight of too limit curve (s). Tre liest cusess seclude the pressure dif tssentsat atteses tne pelat of sistee pressuse seasusesset ans the pressure sa the seacter vessel regisa centeelling the limit j
200 turve. Inag sacogde an aedelsenal easgos of safety les possible anstsueeat estes, (25 psi
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Em UNIT 3 i REACTOR COOLANT SYSTEM INSERVICE LEAK AND llVDR0 STATIC TEST AND C00LDOWN LIMITATIONS APPLICABLE FOR FIRST 8.0 EFPY i
OCONEE NUCLEAR STATION f1Ture 3.1.2-3C
' 3.1.3
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Minimum conditiono for criticality
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' Specification 3.1.3.1 The reactor coolant temtarature shall be above 5250F except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply.
3.1.3.2 Reactor coolant temperature shall be above the criticality limit of 3.1.2-1A (Unit 1) 3.1.2-1B (Unit 2) 3.1.2-lc (Unit 3) l 3.1.3.3 When the reactor coolant temperature is below the minimum tem-perature specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply, the reactor shall be suberitical by an amount equal to or greater than the calculated reactivity insertion due to depressurization.
3.1.3.4 The reactor shall be maintained suberitical by at least 1%Ak/k until a steam bubble is formed and a water level between 80 and 396 inches is established in the pressurizer.
3.1.3.5 Except for physics tests and as limited by 3.5.2.1, safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality. The regulating rods shall then be positioned within their position limits defined by Specification 3.5.2.5 prior to deboration.
Bases l
At the beginning of the initial fuel cycle, the moderator temperature coeffi-cient is expected to be slightly positive at operating temperatures with the operating configuration of control rods.(1) Calculations show that above 5250F, the consequences are acceptable.
Since the moderator temperature coefficient at lower temperatures will be I
less negative or more positive than at operating temperature,(2) startup and l operation of the reactor when reactor coolant temperature is less than 5250F is prohibited except where necessary for low power physics rests.
The potential reactivity insertion due to the moderator pressure coefficient (2) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.lAk/k.
During physics tests, special operating precautions will be taken. In addi-tion, the strong negative Doppler coefficient (1) and the small integrated ak/k would limit the magnitude of a power excursion resulting from a reduc-tion of moderator density.
The requirement that the reactor is not to be made critical below the limits of Specification 3.1.2.1 provides increased assurance that the proper rela-l tionship between primary coolant pressure and temperature will be maintained relative to the NDTT of the primary coolant system. Heatup to this tempera-l ture will be accomplished by operating the reactor coolant pumps.
3.1-8
3 ,
If,the chutdown margin rcquircd by Sp:cificction 3.5.2 in maintnined, thera
'is no possibility of an accidental criticality as a result of a decrease of coolant pressure.
The requirement for pressurizar bubble forastion and specified water level when the reactor is less than 1% suberitical will assure that the reactor coolant afates cannot become solid in the event of a rod withdrawal accident or a startup accident.(3)
The requirement that the safety rod groups be fully withdrawn before crici-cality ensures shutdown capability during startup. This does not prohibit rod latch confirmation, i.e. , withdrawal by group to a maximum of 3 inches withdrawn of all seven groups prior to safety rod withdrawa1. ,
The requirement for regulating rods being within their rod position limits ensures that the shutdown margin and ejected rod criteria at hoe zero power are not violated.
REFERENCES (1) FSAR, Section 3 (2) FSAR, Section 3.2.2.1.4 (3) FSAR, Supplement 3 Answer 14.4.1 l
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3.1-9 1
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