Proposed Tech Spec Changes Re Core Protection Safety Limits, Protective Sys Max Allowable Setpoints & Rod Position LimitsML15223A893 |
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Oconee  |
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Issue date: |
05/19/1983 |
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DUKE POWER CO. |
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ML15223A892 |
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NUDOCS 8305270279 |
Download: ML15223A893 (17) |
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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) 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
Attachment 1 Duke Power Company Oconee Nuclear Station Proposed Technical Specification Revision Oconee 1 Cycle 8 Pages Removed 2.1-2 2.1-3 2.1-7 2.3-8 3.5-9 3.5-10 3.5-15 3.5-15a 3.5-15b 3.5-18 3.5-18a 3.5-18b 3.5-18c 3.5-18d 3.5-18e 3.5-21 3.5-21a 3.5-21b 3.5-24 3.5-24a 3.5-24b
'8305270279 830519 PDR ADOCK 05000269 P
PDR
can be related to DNB through the use of the BAW-2 correlation (1).
The BAW-2 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. A DNBR of 1.30 corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is.actually measured.
The curve presented in Figure 2.1-1A represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 106.5 percent of 131.3 x 106 lbs/hr).
This curve is based on the combination of nuclear power peaking factors, with potential effects of fuel densification and rod bowing, which result in a more conservative DNBR than any other shape that exists during normal operation.
The curves of Figure 2.1-2A are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and rod bowing:
- 1.
The 1.30 DNBR limit produced by the combination of the radial peak, axial peak and position of the axial peak that yields no less than a 1.30 DNBR.
- 2.
The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.5 kw/ft for 8C, 9 and 10 Batches of fuel and 17.6 kw/ft for the 10A, 10B gadolinia fuel Batch for Unit 1.
Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power imbalance produced by the power peaking.
The specified flow rates of Figure 2.1-3A correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.
The curve of Figure 2.1-1A is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3A.
The magnitude of the rod bow penalty applied to each fuel cycle is equal to or greater than the necessary burnup dependent DNBR rod bow penalty for the ap plicable cycle minus a credit of 1% for the flow area reduction factor used in the hot channel analysis. All plant operating limits are based on a minimum DNBR criteria of 1.30 plus the amount necessary to offset the reduction in DNBR due to fuel rod bow. (3) 2.1-2
The maximum thermal power for three-pump opertion is 89.899 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.07 =
79.929 percent power plus the maximum calibration and instrument error. The maximum thermal power for other coolant pump conditions is produced in a similar manner.
For each curve of Figure 2.1-3A a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pump situation. The curve of Figure 2.1-1A is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3A.
References (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March, 1970.
(2) Oconee 1, Cycle 4 - Reload Report - BAW-1447, March, 1977.
(3) Oconee 1, Cycle 8 - Reload Report - BAW-1774, February, 1983.
2.1-3
Thermal Power Level, S 120 1 = 1.571
(-41.0,112.0)
(33. 0,112)
ACCEPTABLE 110 2
(-48.0, 101.0) 4 PUMP OPERATION 100 (33(8.0 095.5)99
(-41.0, 89.899) g 3.,9 89 a ACCEPTABLE S443 PUMP
(-48.0,78.899)
OPAP 80 OPERATION (48.0,73.399) 70 I
(4.0, 62.713)
(33. 0, 62.73)
ACCEPTALE 6
4 O.EAT.2N (48. 0, 51.73) 4,312 PUMP 50 OPERATION(48.
20 30I UNACCEPTABLE
- 1 UNACCEPTABLE OPERATION O
0PERATION I
10
-S0
-40
-20 0
20 40 so Reactor Power Imoalance, CORE PROTECTION SAFETY LIMITS, UNIT 1 UKEPOWi OCONEE NUCLEAR STATION Figure 2.1-2A 2.1-7
Tnermal Power Level,
(-17.0, 107.0)0,107.0 a1,
- 1. 0625 ACCEPTALE 3750 4 PUMP
(-33.0,90.0)
OPERATION I.77.2 17, 79.-92 (33.-0,85.0)
(0017779.92)2 UNACCEPTABLE UNACCEPTABLE OPERATION ACCEPTAI OPERATION 4,3 PUMP
(-33.0, 2.92)
OPERATION
(-17, 1 60 (305.2 17,52. 43)
(17.
2.43)
I ACCEPTABLE 40 14, 3,2 PUmP
(-33.0,35.43)
OPE RIaP OPERATION I(33.0,30.43) 1 20
.60 1401
-20 0
20 40 Reactor Power Imbalance, i PROTECTIVE SYSTEM1 iMXIMUM ALLOWABLE SETFOINTS, UNIT 1 OCONEE NUCLEAR STATION Figure 2.3-2A 2.3-8
. If the maximum positive quadrant power tilt exceeds the Maximum Limit of Table 3.5-1, the reactor shall be shut down within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the ther mal power and the Nuclear Overpower Trip Setpoints allowable for the reactor coolant pump combination are restricted by a reduc tion of 2% of thermal power for each 1% tilt for the maximum tilt observed prior to shutdown.
- g.
Quadrant power tilt shall be monitored on a minimum frequency of once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 15% full power.
3.5.2.5 Control Rod Positions
- a.
Technical Specification 3.1.3.5 does not prohibit the exercising of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.
- b.
Except for physics tests, operating rod group overlap shall be 25% t 5% between two sequential groups.
If this limit is ex ceeded, corrective measures shall be taken immediately to achieve an acceptable overlap. Acceptable overlap shall be attained within two hours or the reactor shall.be placed in a hot shutdown condition within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c.
Position limits are specified for regulating and axial power shaping control rods. Except for physics tests or exercising control rods, the regulating control rod insertion/withdrawal limits are specified on figures 3.5.2-lAl and 3.5.2-1A2, (Unit 1);
3.5.2-IB1, 3.5.2-1B2, and 3.5.2-1B3 (Unit 2); 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3 (Unit 3) for four pump operation, on figures 3.5.2-2A1 and 3.5.2-2A2, (Unit 1); 3.5.2-2B1, 3.5.2-2B2, and 3.5.2-2B3 (Unit 2); figures 3.5.2-2C1, 3.5.2-2C2, and 3.5.2-2C3 (Unit 3) for three pump operation, and on figures 3.5.2-2A3 and 3.5.2-2A4, (Unit 1); 3.5.2-2B4, 3.5.2-2B5, and 3.5.2-2B6 (Unit 2); figures 3.5.2-2C4, 3.5.2-2C5, and 3.5.2-2C6 (Unit 3) for two pump operation. Also, excepting physics tests or exercising control rods, the axial power shaping control rod insertion/withdrawal limits are specified on figures 3.5.2-4A1, and 3.5.2-4A2, (Unit 1); 3.5.2-4B1, 3.5.2-4B2, and 3.5.2-4B3 (Unit 2); 3.5.2-4C1, 3.5.2-4C2, and 3.5.2-4C3 (Unit 3).
If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. An acceptable control rod position shall then be attained within two hours.
The minimum shutdown margin required by Specification 3.5.2.1 shall be maintained at all times.
3.5-9
3.5.2.6 Xenon Reactivity Except for physics tests, reactor power shall not be increased above the power level-cutoff shown in Figures 3.5.2-lAl, 3.5.2-1A2, for Unit 1; Figures 3.5.2-IBI, 3.5.2-1B2, and 3.5.2-1B3, for Unit 2; and Figures 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3 for Unit 3 unless one of the following conditions is satisfied:
- 1.
Xenon reactivity did not deviate more than 10 percent from the equilibrium value for operation at steady state power.
- 2.
Xenon reactivity deviated more than 10 percent but is now within 10 percent of the equilibrium value for operation at steady state rated power and has passed its final maximum or minimum peak during its approach to its equilibrium value for operation at the power level cutoff.
- 3.
Except for xenon free startup (when 2. applies), the reactor has operated within a range of 87 to 92 percent of rated thermal power for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
3.5.2.7 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.
Except for physics tests, imbalance shall be maintained within the envelope defined by Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.3-3B1, 3.5.3-3B2, 3.5.2-3B3, 3.5.2-3C1, 3.5.2-3C2, and 3.5.2-3C3.
If the imbalance is not within the envelope defined by these figures, cor rective measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalance is not achieved within two hours, reactor power shall be reduced until imbalance limits are met.
3.5.2.8 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the manager or his designated alternate.
3.5.2.9 The operational limit curves of Technical Specifications 3.5.2.5.c and 3.5.2.7 are valid for a nominal design cycle length, as defined in the Safety Evaluation Report for the appropriate unit and cycle.
Operational beyond the nominal design cycle length is permitted pro vided that an evaluation is performed to verify that the operational limit curves are valid for extended operation. If the operational limit curves are not valid for the extended period of the operation, appropriate limits will be established and the Technical Specification curves will be modified as required.
3.5-10
(204, 102)
(280, 102) 0 OPERATION
NOT ALLOWED SHUTDOWN MARGIN (275 92)
POWER LEVEL LIMIT CUTOFF
= 100%FP gg8 (273,80)
OPERATION RESTRICTED 60 139,50)
(200,50) 40 OPERATION ACCEPTABLE 2
20 C
(0.7. 5) 0 f
0 /
100 200 300 Roa Inaax, s i tnarawn GR 5t 0
75 100 GR 6 L
0 25 75 100 GR 7 I
I 0
25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 0 TO 50 (+10, -0) EFPD, UNIT 1 UKEPOWER OCONEE NUCLEAR STATION Figure 3.5.2-1Al 3.5-15
(238, 102) 275. 102) 100 OPERATION 300,102)
POWER NOT ALLOWED (275, 92)
LEVEL SHUTDOWN CUTOFF MARGIN 100% FP LINIT (266,80)
S 60 (173,50)
(200,50) 40 OPERATION ACCEPTABLE 20 (0,
- 6. 8
( 1 5 0
I 0
100 200 300 Roo Inaex, Itn arawn GR 5' 0
75 100 GR S 0 25 75 100 GR 7 t 0
25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION AFTER 50
(+10, -0) EFPD, UNIT 1 ne'~owE' OCONEE NUCLEAR STATION Figure 3.5.2-lA2 3.5-15a
100 OPERATION NOT ALLOWED so SHUTDOWN (204,77)
(265,77)
NARG*
(3()0,77 )
HARGIN 40 (139, 38)
OPERATION ACCEPTABLE 20 0- (0, 6.1)
A 0
100 200 300 Roa Inaex, S Witnarawn SR 5I 0
75 100
- R 6 IL 0
25 75 100 GR 7 I
-I 0
25 100 ROD POSITION LIMITS FOR THREE PUMP OPERATION FROM 0 TO 50
(+10, -0) EFPD, UNIT 1 0 0KEDwni OCONEE NUCLEAR STATION Figure 3.5.2-2Al 3.5-18
100 OPERATION RESTRICTED so OPERATION (238.77) 259.4,77)
NOT ALLOWED 300,77 SHUTDOWN MARGIN so
.LIMIT (200, 50) 40 (173.38)
OPERATION 20 ACCEPTABLE 0(0 5 6) 0 100 200 300 Roo Inoex, s Witnarawn SR 5 1
0 75 100 GR 8 0
25 75 100 CR 7 t
0 25 100 ROD POSITION LIMITS FOR THREE PUMP OPERATION AFTER 50 (+10, -0) EFPD, UNIT 1 oInEP OCONEE NUCLEAR STATION Figure 3.5.2-2A2 3.5-18a
log 100 80
- OPERATION NOT ALLOWED S 60 SHUTDOWN (0,2 MARGIN LIMIT 40 20 (139,26)
OPERATION ACCEPTABLE (0,4.8 (64,8).
0 100 200 300 Roa Inaex, W
Witnarawn GR 5 1 0
75 100 GR 6 t
I I
I 0
25 75 100 GR7 0
25 100 ROD POSITION LIM4ITS FOR TWO PUMP OPERATION FROM 0 TO 50 (+10, -0) EFPD, UNIT 1 UKEPowEr OCONEE NUCLEAR STATION Figure 3.5.2-2A3 3.5-18b
80 80OPERATION NOT ALLOWED 60 (28,2 SHUTD0WN(28i)
MARGIN (300,52)
LIMIT 40
'3 OPERATION ACCEPTABLE 20 (0
,44) 100 200 300 Ro Index, i Ii triarawn GR 5' 0
75 100 GR 5' 0
25 75 100 GR 7 1 0
25 100 ROD POSITION LIMITS FOR TWO PUMP OPERATION AFTER 50
(+10, -0) EFPD, UNIT 1 3.5-18c OW OCONEE NUCLEAR STATION Figure 3.5.2 A4
OPERATION RESTRICTED 7,02)
(20, 102) 100 25,92) 90 (25, 92) 80 30,80 70 OPERATION ACCEPTABLE 60 50 40 W
30 20 10
-40
-30
-20
-10 0
10 20 30 40 Axial Power Imoalance, S POWER IMBALANCE LIMITS FOR OPERATION FROM 0 TO 50 (+10, -0)
EFPD, UNIT 1 DuxtpowER OCONEE NUCLEAR STATION Figure 3..5.2-3A1 3.5-21
OPERATION RESTRICTED (20, 102) 00 (3052)
(* 29, 92) 90 OPERATION 70 ACCEPTABLE 50 40 30 10
-40
-30
-20 010 10 20 30 40 Axial Power Imoalance, S POWER IMBALANCE LIMITS FOR OPERATION AFTER 50 (+10, -0)
EFPD, UNIT 1 opuowER OCONEE NUCLEAR STATION 3.5-21a Figure 3.5.2-3A2
mem 3a a'6 Qca af c
cm a a CD C2 sa Jaa*
18 J
U 3%4" APSR POSITION LIMITS FOR OPERATION FROM 0 TO 50
(+10,
-0)
- EFPD, UNIT 1
3DIJEEPOWE OCONEE NUCLEAR STATION Figure 3.5.2-4A1
wa caa s
-c
=
6h=
664
- r.
caa m
p-c n
3.5 2 4 F i u
e 3.
A