Proposed TS 3.7-1 Reducing Maximum Allowable Linear Power level-high Trip Setpoints Associated W/Inoperable Steam Line Safety ValvesML20044D038 |
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Arkansas Nuclear ![Entergy icon.png](/w/images/7/79/Entergy_icon.png) |
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Issue date: |
05/07/1993 |
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ENTERGY OPERATIONS, INC. |
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ML20044D037 |
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References |
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NUDOCS 9305170100 |
Download: ML20044D038 (11) |
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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20216G0881999-09-19019 September 1999 Proposed Tech Specs Pages,Changing TS 4.18.5.a.9 & Associated Bases for One Cycle of Operation Through End of Cycle 16 for Suppl Use of OTSG Repair Roll Technology ML20216G0721999-09-17017 September 1999 Proposed Tech Specs,Revising Once Through SG Tubing Surveillance Requirements to Provide Alternate Repair Criteria for Volumetric Integranular Attack ML20212F5851999-09-17017 September 1999 Proposed Tech Specs Pages,Changing TS by Revising Curie Limits for Radioactive Gas Storage Tanks ML20211P0061999-09-0707 September 1999 Proposed TS in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria ML20211B3291999-08-19019 August 1999 Proposed TS 4.18.3.a.4 Clarification Supporting Tube End Cracking Alternative Repair Criteria ML20211C1701999-08-18018 August 1999 Proposed Tech Specs Revising ESAS Low RCS Pressure Setpoint to Ensure That TS Adequately Bounds Safety Analysis ML20211C4101999-08-18018 August 1999 Proposed Tech Specs Removing Requiremets for Tube Repair (Sleeving) & Clarifying Requirements for Preservice & ISI of Tubes to Facilitate Replacement of Original SGs ML20210K5121999-07-29029 July 1999 Proposed Tech Specs,Revising Requirements Associated with Plant Station Batteries & DC Sources to 125 Vdc Switchyard Distribution Sys ML20210K3571999-07-29029 July 1999 Proposed Tech Specs Revising SRs Pertaining to ISI Requirements for SG Tube Circumferential Cracking at top-of-tubesheet During 2P99 Planned mid-cycle Outage ML20209H5811999-07-14014 July 1999 Proposed Tech Specs Revising RB Structural Integrity Requirements Re ISI Reporting for Containment Structures, Tendons & Anchorages ML20209H8701999-07-14014 July 1999 Proposed Tech Specs 6.2.2 & 6.2.2.h,replacing Functional Description of Operations Staff Middle Manager with Ref to Functional Description Assistant Operations Manager & Revising TS Table 6.2-1 Item 5 & TS 6.2.2.f ML20209H9721999-07-14014 July 1999 Proposed Tech Specs Changes for Elimination of Programmatic Requirements Associated with PASS & Regulatory Relief from PASS Requirements of NUREG-0737 & RG 1.97 ML20195D2371999-06-0101 June 1999 Proposed Tech Specs Re Surveillance Requirements & Applicable Bases Relevant to ISI Requirements for Portions of once-through SG (OTSG) Tubes Adjacent to Primary Cladding Region of tube-to Tubesheet Roll ML20206T2651999-05-14014 May 1999 Proposed Tech Specs Revising SG Tubing SRs ML20206Q1281999-05-10010 May 1999 Proposed Tech Specs Bases,Page B 2-2,discussing Guidance to Be Used When Applying Later Editions & Addendas of ASME Code to RCS Components ML20205P4081999-04-0909 April 1999 Proposed Tech Specs Changes,Revising Reactor Bldg Structural Integrity Requirements ML20205P6291999-04-0909 April 1999 Proposed Tech Specs Changes,Revising Battery Requirements ML20205P3731999-04-0909 April 1999 Revised Proposed TS Page 5-1 Re Design Features Conversion ML20205P6931999-04-0909 April 1999 Proposed Tech Specs Change,Adding Specification 3.0.6 & Associated Bases to ANO-2 ML20207C0631999-02-25025 February 1999 Proposed Tech Specs Change Revising Action 2 Associated with Table 3.3-1, Reactor Protective Instrumentation. Change Adds Footnote to Action 2 Allowing Startup with Functional Units Associated with Channel D Excore Inoperable ML20207M7561999-02-24024 February 1999 Proposed Tech Specs,Allowing Startup & Continued Operation of Plant with Channel D Detector Until New Detector Is Installed During mid-cycle Outage Scheduled for Nov 1999 ML20198K1341998-12-17017 December 1998 Corrected Tech Specs Pages to 980629 Application for Amend to License NPF-6 ML20198A3661998-12-0808 December 1998 Proposed Tech Specs B3/4 6-3 & B3/4 6-4,deleting All Requirements Associated with Sodium H Addition Sys & Adding New Requirements for Trisodium Phosphate to Be Stored in Containers Located on Floor of Cb ML20196B4381998-11-24024 November 1998 Proposed Tech Specs Pages,Changing Administrative Controls Section to Implement Consolidated QA Plan Manual & Clarifying Responsibilities on STA Position on Shift ML20153D8491998-09-17017 September 1998 Proposed Tech Specs 2.2.1,3.3.1.1 & 3.3.2.1 Re Operating Bypass Setpoint ML20236Y5081998-08-0606 August 1998 Proposed Tech Specs Section 3.3.4.B Re Sodium Hydroxide Tank Limits ML20236Y5251998-08-0606 August 1998 Proposed Tech Specs Section 3.1.3.2,revising Action Requirements for CEA Position Indicator Channels ML20236Y5951998-08-0606 August 1998 Proposed Tech Specs,Revising Requirements Associated W/Units 1 & 2 Control Room Emergency Ventilation Sys ML20236Q2781998-07-13013 July 1998 Proposed Tech Specs Re Safety Limits & Limiting Safety Settings ML20236G6921998-06-29029 June 1998 Proposed Tech Specs,Revising TS 3.4.2 & TS 3.4.12 Re Reactor Coolant Sys Safety Valves & Low Temp Overpressure Protection ML20236G5421998-06-29029 June 1998 Proposed Tech Specs,Revising Surveillance Testing Requirements for Unit 2 Direct Current Electrical Distribution Sys ML20236G4911998-06-29029 June 1998 Proposed Tech Specs Re MSSVs & Pressurizer Safety Valves ML20236G5631998-06-29029 June 1998 Proposed Tech Specs,Revising Table 3.3-4 to Provide Range of Acceptable Values for 4160 Volt Bus Loss of Voltage Values ML20217M2621998-04-30030 April 1998 Proposed Tech Specs Re DG Load Rejection Testing ML20217N6141998-04-30030 April 1998 Proposed Tech Specs 1.6.1,revising Definition of Quadrant Power Tilt ML20217M4871998-04-28028 April 1998 Proposed Tech Specs Changing Order of Channel Rotation During Outage for Scheduling Flexibility,While Still Requiring Rotation Schedule to Be Established for Operating Cycle ML20217P8141998-04-0101 April 1998 Proposed Tech Specs Re Upper Tubesheet Volumetric Indications to Remain in Svc for Cycle 15 ML20216J7841998-03-17017 March 1998 Proposed Tech Specs Bases Change Concerning Plant Protection Sys Matrix Logic Testing ML20203D1061998-02-18018 February 1998 Proposed Tech Specs Revising Surveillance Testing Requirements for Emergency Feedwater Sys ML20202G9771998-02-0909 February 1998 Proposed Tech Specs Pages,Allowing Use of Repair Roll Technology (Reroll) for Upper Tubesheet Region of SGs ML20197A8421997-12-12012 December 1997 Proposed Tech Specs Re Establishment of Alternate Repair Criteria for Outer Diameter Intergranular Attack in Upper Tubesheet of Unit 1 SGs ML20211C6441997-09-23023 September 1997 Proposed Tech Specs,Modifying RPS & ESFAS Trip Setpoint & Allowable Values for SG Low Pressure ML20211C8911997-09-23023 September 1997 Proposed Tech Specs Reducing Min Required Reactor Coolant Sys Flow Rate Requirements Until ANO-2 SGs Are Replaced ML20151J3371997-07-28028 July 1997 Proposed Tech Specs,Modifying Actions Associated W/Ts Table 3.3-1 for Reactor Protective Instrumentation & Table 3.3-3 for Esfa Sys Instrumentation ML20149K0551997-07-21021 July 1997 Proposed Tech Specs,Revising Surveillance Testing Requirements for Facility EFW Sys as Specified in Specification 4.7.1.2 ML20141C5641997-06-18018 June 1997 Proposed Tech Specs Changes Restoring Table 2 to Table II & Revision to One of Offsite Dose Calculation Manual, Pages Previously Submitted.Also Includes Amended Pages Modified to Amends Issued Since Change Request Submitted ML20141A8001997-05-0808 May 1997 Proposed Tech Specs 3/4.3.1 & 3/4.3.2 Requesting Removal of Reference to CE NPSD-576 ML20138H8521997-05-0202 May 1997 Proposed Tech Specs,Submitting Rev to Original TS Change Request to Reduce Limit Specified in Section 3.1.6.3.b for Leakage Through SG Tubes of Any One Sg,Per SG Tube Insp Surveillance Requirements ML20138D6411997-04-24024 April 1997 Proposed Tech Specs Revising Requirements for Sample Scope Initial Size & Expansion for Inservice Insp of Steam Generator Tube Sleeves ML20141H2621997-04-18018 April 1997 Proposed Tech Specs,Permitting Delay of Scheduled three-yr Inspections of Flywheels on All Four RCPs Until Completion of 2R13 Refueling Outage,Scheduled for Winter of 1998-99 1999-09-07
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20217N3171999-10-19019 October 1999 Rev 1,change 024-01-0 to OP-1903.011, Emergency Response Notifications ML20217J7641999-10-0505 October 1999 Rev 1 to Unit 1 Technical Requirement Manual, Replacing Index Page ML20216G0881999-09-19019 September 1999 Proposed Tech Specs Pages,Changing TS 4.18.5.a.9 & Associated Bases for One Cycle of Operation Through End of Cycle 16 for Suppl Use of OTSG Repair Roll Technology ML20212F5851999-09-17017 September 1999 Proposed Tech Specs Pages,Changing TS by Revising Curie Limits for Radioactive Gas Storage Tanks ML20216G0721999-09-17017 September 1999 Proposed Tech Specs,Revising Once Through SG Tubing Surveillance Requirements to Provide Alternate Repair Criteria for Volumetric Integranular Attack ML20211P0061999-09-0707 September 1999 Proposed TS in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria ML20211B3291999-08-19019 August 1999 Proposed TS 4.18.3.a.4 Clarification Supporting Tube End Cracking Alternative Repair Criteria ML20211C4101999-08-18018 August 1999 Proposed Tech Specs Removing Requiremets for Tube Repair (Sleeving) & Clarifying Requirements for Preservice & ISI of Tubes to Facilitate Replacement of Original SGs ML20211C1701999-08-18018 August 1999 Proposed Tech Specs Revising ESAS Low RCS Pressure Setpoint to Ensure That TS Adequately Bounds Safety Analysis ML20210N5541999-08-0404 August 1999 Rev 28,Permanent Change 4 to OP-1903.061, Communication Equipment Test ML20210N5571999-08-0404 August 1999 Rev 23,Permanent Change 5 to OP-1903.011, Emergency Response/Notifications ML20210K5121999-07-29029 July 1999 Proposed Tech Specs,Revising Requirements Associated with Plant Station Batteries & DC Sources to 125 Vdc Switchyard Distribution Sys ML20210K3571999-07-29029 July 1999 Proposed Tech Specs Revising SRs Pertaining to ISI Requirements for SG Tube Circumferential Cracking at top-of-tubesheet During 2P99 Planned mid-cycle Outage ML20209H5811999-07-14014 July 1999 Proposed Tech Specs Revising RB Structural Integrity Requirements Re ISI Reporting for Containment Structures, Tendons & Anchorages ML20209H8701999-07-14014 July 1999 Proposed Tech Specs 6.2.2 & 6.2.2.h,replacing Functional Description of Operations Staff Middle Manager with Ref to Functional Description Assistant Operations Manager & Revising TS Table 6.2-1 Item 5 & TS 6.2.2.f ML20209H9721999-07-14014 July 1999 Proposed Tech Specs Changes for Elimination of Programmatic Requirements Associated with PASS & Regulatory Relief from PASS Requirements of NUREG-0737 & RG 1.97 ML20195D2371999-06-0101 June 1999 Proposed Tech Specs Re Surveillance Requirements & Applicable Bases Relevant to ISI Requirements for Portions of once-through SG (OTSG) Tubes Adjacent to Primary Cladding Region of tube-to Tubesheet Roll ML20206T2651999-05-14014 May 1999 Proposed Tech Specs Revising SG Tubing SRs ML20206Q1281999-05-10010 May 1999 Proposed Tech Specs Bases,Page B 2-2,discussing Guidance to Be Used When Applying Later Editions & Addendas of ASME Code to RCS Components ML20206J8021999-05-0404 May 1999 PC-03 to Procedure/Work Plan 1904-002, Offsite Dose Projections Rdacs Computer Method ML20205P6931999-04-0909 April 1999 Proposed Tech Specs Change,Adding Specification 3.0.6 & Associated Bases to ANO-2 ML20205P3731999-04-0909 April 1999 Revised Proposed TS Page 5-1 Re Design Features Conversion ML20205P6431999-04-0909 April 1999 Revised Proposed Technical Requirement Manual Pages ML20205P4081999-04-0909 April 1999 Proposed Tech Specs Changes,Revising Reactor Bldg Structural Integrity Requirements ML20205P6291999-04-0909 April 1999 Proposed Tech Specs Changes,Revising Battery Requirements ML20207C0631999-02-25025 February 1999 Proposed Tech Specs Change Revising Action 2 Associated with Table 3.3-1, Reactor Protective Instrumentation. Change Adds Footnote to Action 2 Allowing Startup with Functional Units Associated with Channel D Excore Inoperable ML20207M7561999-02-24024 February 1999 Proposed Tech Specs,Allowing Startup & Continued Operation of Plant with Channel D Detector Until New Detector Is Installed During mid-cycle Outage Scheduled for Nov 1999 ML20198K1341998-12-17017 December 1998 Corrected Tech Specs Pages to 980629 Application for Amend to License NPF-6 ML20198A3661998-12-0808 December 1998 Proposed Tech Specs B3/4 6-3 & B3/4 6-4,deleting All Requirements Associated with Sodium H Addition Sys & Adding New Requirements for Trisodium Phosphate to Be Stored in Containers Located on Floor of Cb ML20196B4381998-11-24024 November 1998 Proposed Tech Specs Pages,Changing Administrative Controls Section to Implement Consolidated QA Plan Manual & Clarifying Responsibilities on STA Position on Shift ML20196C9761998-11-24024 November 1998 Rev 1 to Unit 2 Technical Requirements Manual ML20154D6681998-10-0101 October 1998 Rev 28 to Procedure/Work Plan 1904.002, Offsite Dose Projections Rdacs Computer Method ML20153D8491998-09-17017 September 1998 Proposed Tech Specs 2.2.1,3.3.1.1 & 3.3.2.1 Re Operating Bypass Setpoint ML20153H1651998-08-12012 August 1998 Rev 0 to ANO Units 1 & 2 Technical Requirements Manual ML20236Y5251998-08-0606 August 1998 Proposed Tech Specs Section 3.1.3.2,revising Action Requirements for CEA Position Indicator Channels ML20236Y5081998-08-0606 August 1998 Proposed Tech Specs Section 3.3.4.B Re Sodium Hydroxide Tank Limits ML20236Y5951998-08-0606 August 1998 Proposed Tech Specs,Revising Requirements Associated W/Units 1 & 2 Control Room Emergency Ventilation Sys ML20236Q2781998-07-13013 July 1998 Proposed Tech Specs Re Safety Limits & Limiting Safety Settings ML20236G6921998-06-29029 June 1998 Proposed Tech Specs,Revising TS 3.4.2 & TS 3.4.12 Re Reactor Coolant Sys Safety Valves & Low Temp Overpressure Protection ML20236G5421998-06-29029 June 1998 Proposed Tech Specs,Revising Surveillance Testing Requirements for Unit 2 Direct Current Electrical Distribution Sys ML20236G5631998-06-29029 June 1998 Proposed Tech Specs,Revising Table 3.3-4 to Provide Range of Acceptable Values for 4160 Volt Bus Loss of Voltage Values ML20236G4911998-06-29029 June 1998 Proposed Tech Specs Re MSSVs & Pressurizer Safety Valves ML20199F7211998-05-22022 May 1998 Rev 8 to Simulator Mod Control,Training Desk Guide 6.3 ML20217N6141998-04-30030 April 1998 Proposed Tech Specs 1.6.1,revising Definition of Quadrant Power Tilt ML20217M2621998-04-30030 April 1998 Proposed Tech Specs Re DG Load Rejection Testing ML20217M4871998-04-28028 April 1998 Proposed Tech Specs Changing Order of Channel Rotation During Outage for Scheduling Flexibility,While Still Requiring Rotation Schedule to Be Established for Operating Cycle ML20217P8141998-04-0101 April 1998 Proposed Tech Specs Re Upper Tubesheet Volumetric Indications to Remain in Svc for Cycle 15 ML20216J7841998-03-17017 March 1998 Proposed Tech Specs Bases Change Concerning Plant Protection Sys Matrix Logic Testing ML20203D1061998-02-18018 February 1998 Proposed Tech Specs Revising Surveillance Testing Requirements for Emergency Feedwater Sys ML20202G9771998-02-0909 February 1998 Proposed Tech Specs Pages,Allowing Use of Repair Roll Technology (Reroll) for Upper Tubesheet Region of SGs 1999-09-07
[Table view] |
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l PROPOSED TECHNICAL SPECIFICATION CHANGES :
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3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE "
SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE with lift settings as specified in Table 3.7-5.
APPLICABILITY: MODES 1, 2 and 3*.
ACTION:
a.
With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Linear Power Level-High trip setpoint is reduced per Table 3.7-1; otherwise, be in at Isast HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. -
- b. With one reactor coolant loop and associated steam generator in operation and with one or more main steam line code safety valves associated with the operating steam generator inoperable, operation in MODES 1, 2 and 3 may proceed provided:
- 1. That at least 2 main steam line code safety valves on the non-operating steam generator are OPERABLE, and
- 2. That within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Linear Power Level-High trip setpoint is reduced per Table 3.7-2; otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REOUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.
code safety valves may be gagged and two (one on each header) may be reset for '
the duration of the test to allow the required pressure for the test to be attained. The Reactor Trip Breakers shall be open for the duration of the test.
ARKANSAS - UNIT 2 3/4 7-1 Amendment No. E), 134 i
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$2 .
gj TABLE 3.7-1 3 -
( MAXIMUM ALLOWABLE LINEAR POWER LEVEL-HIGH TRIP SETPOINT WITil INOPERABLE STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS N
Maximum Allowable Linear Maximum Number of Inoperable Safety Power Level-High Trip Setpoint Valves on Anv Oreratine Steam Generator (Percent of RATED THERMAL POWER) i 1 91.0 2 67.7 3 45.0 M>
Y
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3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY CYCLES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% of its design pressure of 1100 psia during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lif t settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The "as-found" requirements are conservative with respect to Section XI of the ASME Boiler and Pressure Vessel Code, 1986 Edition, and Addenda through 1987. The total relieving capacity for all MSSVs is 14,804,000 lb/hr. The ANO-2-fu11 power (100%) steam flow plus 2% for conservatism is 12,950,000 lb/hr.
Therefore, the total relieving capacity of the MSSVs is 114.32% of the 102% power steam flow through each main steam header. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for removing decay heat.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels. The reactor trip setpoint reductions are derived on the following bases:
~
SP = ' x (114.32%)
where:
SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER
~ ARKANSAS . UNIT 2 B 3/4 7-1 Amendment No. 110
... i l
PLANT SYSTEMS BASES f
i i
114.32% = Total maximum safety valve relieving capacity of 102% power ,
steam flow !
X = Total relieving capacity of all safety valves per steam }
line in Ibs/ hour !
Y =
Total Maximum relieving capacity of the inoperable safety I valve (s) in Ibs/ hour. In each case, the valves with the ,
, greatest relieving capacity were assumed inoperable, j 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEM The OPERABILITY of the emergency feedwater system ensures that the I Reactor Coolant System can be cooled down to less than 350"F from normal operating conditions in the event of a total loss of off-site power. ;
Each emergency feedwater pump is capable of delivering a total feedwater flow of 485 gpm at a pressure of 1100 psig to the entrance of i the steam generators. This capacity is sufficient to ensure that adequate -
feedwater flow is available to remove decay heat and reduce the Reactor '
Coolant System temperature to less than 350'F when the shutdown cooling .
system may be placed into operation. !
3/4.7.1.3 CONDENSATE STORAGE TANK i
The OPERABILITY of the condensate storage tank with the minimum water ,
volume ensures that sufficient water is available to maintain the RCS at, !
110T STANDBY conditions for one hour with steam discharge to atmosphere I with concurrent with total loss of off-site power. The contained water j volume Ifmit includes an allowance for water not usable because of tank i discharge line location or other physical characteristics. I l
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ARKANSAS - UNIT 2 B 3/4 7-2 .
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= ENTERGY ARKANSAS NU N AR ONE CALCUIATION COVER SHEET Calc. No.: @E-occ3-o/
fk-7,4r vnit 2 Category O Ca}c.
Title:
A//w-2 _
f%s 4-ode ra #t r y.E system (s): M5' MS4rls Topic (s): [d.77~
Calc. Typer blI Corponent No(s).: MSV- /m2//cotl Plt Area: Bldg. Elev. -
/60W/oof/tocs/ IOS2//W//o3W Room ~
ua11
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Abstract (Included Purpose /Results): kw> em- oEpordinates:
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/e -h> dah L ,1 a //esel n </h ww / ,wk 'r a H2 Ln.-nA/,
MTS Vs.
WeAalk Som w>lr<J ek ,A 3 , -
Rcv. No.: Verification Method: Design Review:
Alternate Calculation: Qualification Testing:
Pcg:s Revised and/or Added:
Purpose of Revisions %!4'. / TS$X 4' #E a e r
- i v Initiating Documents Resulting Docuaent(s) Key Desi dG '2-92-0W7 Yst4 ke, 3.*7.l.l ' 40-6-a-%gn-l2 Input Docs.
fickt rL Ardi k C4~ 2 ovrz. .
Am:nds Cale(s): Afh Supercedes Calc (s): Me Corputer Software (Ver.): W/
By: hde/ VT'G 7/Emr / d/ Rvw'd: ^'/d / /
Chk'd: lwlSlblsk / 4)L / bok3 hyv'd: Jehus Mrub / V3C / 1-3 43
- (Print Name) (Initials)( at'e) (Print Name) (In'itials)(Date)
Check if Additional Revisions:
~
Purpose:
The purpose of this calculation is to determine the i allowed reactor power levels for varying numbers of inoperable main steam safety valves. These power levels will be used as basis for a change to Technical Specification 3.7.1.1 to ensure i the peak design secondary system pressure limit of 110% of 1100 psia is not exceeded (an ASME code requirement). This !
calculation is being performed since Technical Specification !
3.7.1.1 does not, at this time, adequately restrict power ,
levels with inoperable MSSVs (Ref. 1).
t
References:
1
~
- 1) CR-2-92-0492, 12/04/92.
- 2) Calculation No. 90-E-0096-12, Rev. O, Reload Analysis For Cycle 10, 11/6/92.
- 3) ANO-2 Technical Specifications.
- 4) ANO-2 Safety Analysis Report, Amendment No. 10, Section No. !
10.3.2.1.
Discussion of Problem:
In the event of a loss of load transient, such as a Loss of Condenser Vacuum (LOCV), the steam generated by the RCS must be relieved to prevent overpressurization of the secondary side. To accomplish this there are five (5) secondary system safety relief valves on each main steam header, outside containment, upstream of the Main Steam Isolation Valves (MSIVs). The MSSV design includes staggered setpoints such that only the number of valves needed will lift. The purpose of the MSSVs is to limit secondary system pressure to lower than 110% of the nominal design pressure of 1210 psia (design pressure is 1100 psia, Ref. 4) when passing 100% of. design steam flow. If one or more of the MSSVs is inoperable then !
reactor power may need to be adjusted to prevent the steam produced from the RCS from exceeding the degraded steam relief capacity. If the RCS were to produce more steam than the secondary side could relieve, overpressurization of the secondary side upon load rejection could result.
During a load rejection transient the turbine stop valves close and essentially " bottle up" the steam line (SDBCS is assumed to not be in operation to relieve the steam flow). [
Secondary pressure would then rise to the point at which flow >
$ wo,,y CALCULATION NUMBER ,
~ operatims 93-E-0003-01 ARKANSAS 01/15/9: wc 0 @A> NUCLEAR PAGE I REY. DATE BY 1 4 CHK'D ONE NUMBER 0F
4 l
from the relief valve equaled the steam production rate.
I Therefore, to prevent overpressurization of the secondary side !
with inoperable MSSVs, reactor power must be reduced to ensure :
that the RCS does not produce more steam than can be relieved i by the secondary side. j i
Assumptions:
i
- 1) The Steam Dump Bypass Control System (SDBCS) is assumed to .
not be in operation.
- 2) The two steam lines do not communicate or equalize in !'
pressure. This is the same as assuming that one or both MSIVs ;
are closed (worst case configuration).
- 3) The full power steam flow will be increased 2% for conservatism.
l
- 4) The total steam flow from the secondary side relief valves l is equal to the sum of the rated flows of the individual valves. This is very conservative since the flow from a relief ,
valve exceeds rated at inlet pressures above accumulation. t
- 5) Instrument uncertainty will not be included in the reactor !
power levels developed in this calculation. I
- 6) Steam production / flow from each steam generator is assumed to be 50% of the full power steam flow. This ignores any small differences in steam production rates due to tube plugging and ;
sleeving.
- 7) The MSSVs will be assumed to be inoperable in order _from the valve with the highest rated flow to the valve with the lowest rated flow. '
- 8) An implicit assumption in this calculation methodology is that main feedwater temperature is assumed.to be constant for !
all power levels. This means that the steam production rate i
per unit energy produced is assumed to be constant at all power i levels (constant secondary delta-enthalpy). This is i conservative since main feedwater temperature drops with
~
~
reduced power levels which decreases the amount of steam i produced per unit energy due to the increase in feedwater j enthalpy. i
@ s,% CALCULATION NUMBER .
operanons 93-E-0003-01 ARKANSAS 0 01/15/93 h Tuc- NUCLEAR PAGE REV. DATE- BY CHK'D ONE NUMBER 2 OF
_ ___m ___ _
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Analysis:
The full power (100%) steam flow for ANO-2 is 12.7E06 !
lb/hr (Ref. 2). This will be increased 2% for conservatism ;
(assumption No. 3):
1.02*12.7E06 = 12.95E06 lb/hr l The rated flows for the 10 Main Steam Safety Valves (MSSVs) (Ref. 2) are tabulated below Valve Tag Number Rated Flow (lb/hr) 2PSV-1002/1052 1.438E06 2PSV-1003/1004/1053/1054 1.473E06 ,
2PSV-1005/1006/1055/1056 1.509E06 !
For conservatism the total rated flow on one side of the steam line will be taken to be the total rated flow for a steam line !
or: !
1.438 + 2(1.473) + 2(1.509) = 7.402E06 lb/hr ,
This maximum relief rate of 7.402E06 lb/hr is:
i 7.402/(12.95/2) = 114.32% .j of the 102% power steam flow through each main steam header. :
If 1 MSSV is inoperable (assumed to have the highest rated
- flow) the new relief capacity is:
114.32%*(7.402 - 1.509)/7.402 = 91.0%
With 2 MSSVs inoperable: i 114.32%*(7.402 - (2(1.509))/7.402 = 67.7% ;
With 3 MSSVs inoperable:
114.32%*(7.402 - (2(1.509) + 1.473))/7.402 = 45.0% l t
Operation with less than 2 operable MSSVs is not allowed per reference 3.
i dS ento,yy CALCULATION NUMBER :
operations ;
93-E-0003-01 <
ARKANSAS O 01/15/92 @ @C NUCLEAR PAGE !
REV. DATE BY 3 CHK'D ONE NUMBER OF ___ i 4
conclusion:
The allowed reactor power levels based on the number of inoperable MSSVs on one steam line are tabulated below:
l Number of Allowed Reactor l Inoperable MSSVs Power Level (%) l l
0 100 ;
1 91.0 2 67.7 i 3 45.0 i l
l t
1 i
i Oentergy CALCULATION NUMBER operauons 93-E-0003-01 AltKANSAS O 01/15/9: @v Tat NUCLEAll PAGE I ItEV. DATE 4 4 BY CIIK'D ONE NUMBElt cr 1