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MONTHYEARML20137H3301985-11-25025 November 1985 Informs That Revised License Change Application 113 Will Be Submitted within 60 Days.Deferment of Review of Application Verbally Requested in Jul 1985 Project stage: Other ML20138R5381985-12-27027 December 1985 Proposed Tech Specs 3/4.4.1.2 Re Number of Reactor Coolant Loops to Be Operating at Low Power & in Mode 3.Certificate of Svc Encl Project stage: Other ML20138R5351985-12-27027 December 1985 Rev 1 to License Change Application 113,requesting Amend to License NPF-1,revising Tech Specs 3/4.4.1.1 & 3/4.4.1.2 Re Number of Reactor Coolant Loops Operating at Low Power & in Mode 3 Project stage: Other ML20138R5291985-12-27027 December 1985 Forwards Rev 1 to License Change Application 113,requesting Amend to License NPF-1 Re Number of Operating Reactor Coolant Loops at Low Power & in Mode 3 Project stage: Other ML20207J9171986-12-16016 December 1986 Forwards Amend 122 to License NPF-1 & Safety Evaluation. Amend Modifies Tech Spec Sections 3.4.1.1 & 3.4.1.2 Re Number of Reactor Coolant Loops Required to Be in Operation in Mode 3 & During Low Power Operation Project stage: Approval 1985-12-27
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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20204B7691999-03-18018 March 1999 Proposed Tech Specs 5.6 Re High Radiation Area.Certificate of Svc,Encl ML20207J0831999-02-11011 February 1999 Proposed Tech Specs for Trojan Isfsi,Stating Length of Time Air Pads Can Be Installed & Inflated ML20151X8671998-09-10010 September 1998 Rev 2 to Proposed TS Bases,Changing Name of Cooling Sys to Modular Spent Fuel Pool Cooling & Cleanup Sys & Updates Spent Fuel Pool Heat Up Info to 3 1/2 Years After Plant Shutdown ML20238E8291998-08-27027 August 1998 Proposed Tech Specs,Deleting Number of License Conditions & TS Requirements as Prerequisite to Implementing Requested Amend to Transfer Nuclear Sf from Existing 10CFR50 Licensed Area to Proposed 10CFR72,ISFSI Area ML20199J5801998-01-29029 January 1998 Proposed Tech Specs Pages,Deleting Security Plan Requirements of 10CFR50.54(p) & Part 73 for Trojan Plant After Spent Nuclear Fuel Is Moved to Proposed ISFSI Area ML20134C9901997-01-28028 January 1997 Proposed Tech Specs 5.0 Re Administrative Controls ML20134A9571997-01-16016 January 1997 Proposed Tech Specs LCO 3.1.4 Re Spent Fuel Pool Load Restrictions ML20134B5881997-01-16016 January 1997 Proposed Tech Specs Deleting Paragraph 2.(C)(7) to Allow Loading & Handling of Spent Fuel Casks & Insertion of New License Condition Authorizing Loading of Spent Fuel & Other Matls in Fuel Building ML20134B8631997-01-16016 January 1997 Proposed Tech Specs Re Delete Paragraph 2.(C)(7) to Allow pre-operational Testing & Load Handling of Spent Fuel Transfer & Storage Casks in Trojan Fuel Building ML20135B5071996-11-27027 November 1996 Proposed Tech Specs Reflecting New Fuel Debris Storage Container Design & Storage of non-fuel Bearing Components & Fuel Rod Storage Container in Trojan ISFSI ML20129E7341996-10-23023 October 1996 Proposed Tech Specs Re Spent Fuel Pool Debris Processing ML20059A5601993-10-21021 October 1993 Proposed Tech Specs Modifying Security License Conditions for Clarity ML20127P2121993-01-27027 January 1993 Proposed TS 6.2.2.f Re Composition of Fire Brigade & TS Table 6.2-1 Re Min Shift Crew Composition for Shift Manager & Noncertified Operator ML20127M6391992-11-16016 November 1992 Proposed Tech Specs for Deferral of Unscheduled Steam Generator ISI ML20141M1901992-03-27027 March 1992 Proposed Tech Specs Re Surveillance Requirements for Lco. Response to Request for Addl Info on Subj Encl ML20059L6701990-09-21021 September 1990 Proposed Tech Spec Table 3.6-1, Containment Isolation Valves ML20247C8941989-07-19019 July 1989 Proposed Tech Spec Table 3.6-1, Containment Isolation Valves ML20245C8091989-06-12012 June 1989 Proposed Tech Specs,Changing Orifice Size of Steam Line Safety Valves Shown in Tech Spec Table 4.7-1, Steam Line Safety Valves Per Loop ML20235L9441989-02-10010 February 1989 Proposed Tech Specs,Allowing Mods to Control Auxiliary Fuel Bldg Complex Resulting in Increase in Lateral Shear Forces on Any Story ML20235Q9191989-02-10010 February 1989 Proposed Tech Specs Re Limiting Conditions for Operation & Surveillance Requirements for Containment Integrity for Equipment Hatches ML20154D1781988-09-0909 September 1988 Proposed Tech Specs Revising Pressurizer LPSI Setpoint ML20151T5951988-08-12012 August 1988 Proposed Tech Specs,Revising Surveillance Requirements for ECCS Check Valves ML20151A0381988-07-0808 July 1988 Proposed Tech Specs Re Radioactive Gaseous Effluents & Radioactive Liquid Waste Sampling & Analysis Program ML20151U7171988-04-14014 April 1988 Revised Proposed Tech Specs Re Plant Sys,As Part of Rev 1 to License Change Application 142 ML20148K5971988-03-23023 March 1988 Proposed Revised Tech Specs Re RCS ML20148C2101988-03-18018 March 1988 Proposed Tech Specs,Reflecting Revised Offsite & Facility Organization Charts ML20196J9031988-03-0101 March 1988 Proposed Tech Specs Modifying OL NPF-1,removing 3.5% U-235 Limit on Reactor Fuel Assemblies & Increase Enrichment Limit for New Fuel Storage Racks to 4.5%.Certificate of Svc Encl ML20147E7371988-03-0101 March 1988 Proposed Tech Specs Re Containment Isolation Valves ML20149K1901988-02-17017 February 1988 Proposed Tech Specs,Adding New Instruments for Monitoring Steam Generator Blowdown Sys Liquid Effluent Radioactivity ML20149L6981988-02-15015 February 1988 Proposed Tech Specs Re Offsite & Facility Organization Charts.Util Certificate of Svc Encl ML20149H1121988-02-0404 February 1988 Proposed Tech Specs Revising Operability & Surveillance Requirements for Component Cooling Water Sys ML20147D4561988-01-15015 January 1988 Proposed Tech Specs,Revising Offsite & Facility Organization Charts ML20195H8971987-12-24024 December 1987 Proposed Tech Specs Deleting Sections Re Fire Protection ML20237D6971987-12-18018 December 1987 Proposed Tech Specs,Requiring Measurement of Initial Discriminator Bias Curve for Each Detector & Subsequent Discrimination Bias Curves Obtained,Evaluated & Compared to Initial Curves ML20237D6551987-12-16016 December 1987 Proposed Tech Specs Deleting Provision for Continued Plant Operation in Section 3/4.4.6.2 & Requiring Prompt Unit Placement in Cold Shutdown After Pressure Boundary Leakage Occurrence ML20236W7611987-11-20020 November 1987 Proposed Tech Specs Including Effects of Nuclear Fuel Design Changes ML20236Q2271987-11-16016 November 1987 Proposed Tech Specs,Resolving Control Room Habitability Issues ML20236P4391987-11-13013 November 1987 Proposed Tech Specs,Extending Surveillance Time Period for Verifying Control Rod Insertability During Control Rod Worth & Shutdown Margin Tests to 7 Days.Certificate of Svc Encl ML20236N3861987-11-0909 November 1987 Proposed Tech Spec,Revising Tech Spec 3.9.9 to Resolve Inconsistencies within Tech Specs.W/Certificate of Svc ML20236N5881987-08-0606 August 1987 Proposed Tech Spec Section, Bases Correcting Section for Agreement W/Definition of Pressure Boundary Leakage ML20235W5571987-07-14014 July 1987 Proposed Tech Specs Re Solid Radwaste Sys ML20215K8381987-06-22022 June 1987 Proposed Tech Spec Table 3.3-1,revising Operability Requirements for Reactor Trip Breakers & 4.3-1,revising Surveillance Requirements for Manual Reactor Trip,Reactor Trip Breakers & Bypass Breakers ML20210B6241987-04-29029 April 1987 Proposed Tech Specs Revising Typos & Clarifying Appropriate Process to Be Used for Review & Approval of Temporary Changes to Procedures.Certificate of Svc Encl ML20209D0251987-04-20020 April 1987 Proposed Tech Specs Re RCS Surveillance Requirements ML20206D9261987-04-0303 April 1987 Revised Tech Spec Pages 3/4 8-6 & 3/4 8-7 Re Onsite Power Distribution Sys,Per License Change Application 128 ML20206A6221987-04-0303 April 1987 Revised Tech Spec Page 7,clarifying License Change Application 124 Re power-operated Valves ML20245A2631987-02-26026 February 1987 Proposed Tech Specs,Consisting of Revised Pages for License Change Application 124 & License Change Application 135, Adding Steamline Isolation Valves & Valve CV-8825 Back Onto Table 3.6-1 & Excluding Containment Airlocks from Criteria ML20211P0031987-02-20020 February 1987 Proposed Tech Specs Re Steam Generator Tube Plugging Criteria in Tube Sheet Region ML20211H4021987-02-11011 February 1987 Proposed Tech Specs Increasing Setpoint Tolerance for Pressurizer & Main Steam Safety Valves from 1% to 2% ML20207Q5371987-01-20020 January 1987 Proposed Tech Specs,Revising Surveillance Requirements for Safety Injection Sys Accumulator Isolation Valves 1999-03-18
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20210M7291999-08-0505 August 1999 Replacement Pages to PGE-1078, Trojan Nuclear Plant License Termination Plan. with D&D Modeling Runs for Determination of Screening Dcgls & Survey Area Maps Indicating Preliminary Survey Units ML20207D3951999-06-0101 June 1999 Rev 1 to Trojan Nuclear Plant Reactor Vessel & Internals Removal Project Transportation Safety Plan ML20204B7691999-03-18018 March 1999 Proposed Tech Specs 5.6 Re High Radiation Area.Certificate of Svc,Encl ML20207J3911999-03-10010 March 1999 Trojan Nuclear Plant License Termination Plan ML20207J0831999-02-11011 February 1999 Proposed Tech Specs for Trojan Isfsi,Stating Length of Time Air Pads Can Be Installed & Inflated ML20153D3081998-09-21021 September 1998 Rev 0 to PGE-1077, Tnp Reactor Vessel & Internals Removal Project Transportation Safety Plan ML20151X8671998-09-10010 September 1998 Rev 2 to Proposed TS Bases,Changing Name of Cooling Sys to Modular Spent Fuel Pool Cooling & Cleanup Sys & Updates Spent Fuel Pool Heat Up Info to 3 1/2 Years After Plant Shutdown ML20238E8291998-08-27027 August 1998 Proposed Tech Specs,Deleting Number of License Conditions & TS Requirements as Prerequisite to Implementing Requested Amend to Transfer Nuclear Sf from Existing 10CFR50 Licensed Area to Proposed 10CFR72,ISFSI Area ML20199J5801998-01-29029 January 1998 Proposed Tech Specs Pages,Deleting Security Plan Requirements of 10CFR50.54(p) & Part 73 for Trojan Plant After Spent Nuclear Fuel Is Moved to Proposed ISFSI Area ML20196J5131997-07-29029 July 1997 Rev 0 to Trojan Second Integrated Test Rept,Filter Media Processing Sys Trojan Nuclear Plant Spent Fuel Pool Debris Project ML20196J5241997-07-29029 July 1997 Rev 2 to In-Canal Equipment,Filter Media Processing Sys, Trojan Nuclear Plant Spent Fuel Pool Debris Project ML20137T9361997-04-0808 April 1997 Rev 1 to Trojan Proof of Principle Test Summary ML20137G1731997-03-20020 March 1997 Rev 6 to Trojan Plant Procedure,Tpp 20-2, Radiation Protection Program ML20137K5151997-03-0404 March 1997 Rev 0 to Trojan Proof of Principle Test Summary ML20138L1901997-01-30030 January 1997 Reactor Vessel & Internals Removal Plan ML20134C9901997-01-28028 January 1997 Proposed Tech Specs 5.0 Re Administrative Controls ML20134B8631997-01-16016 January 1997 Proposed Tech Specs Re Delete Paragraph 2.(C)(7) to Allow pre-operational Testing & Load Handling of Spent Fuel Transfer & Storage Casks in Trojan Fuel Building ML20134B5881997-01-16016 January 1997 Proposed Tech Specs Deleting Paragraph 2.(C)(7) to Allow Loading & Handling of Spent Fuel Casks & Insertion of New License Condition Authorizing Loading of Spent Fuel & Other Matls in Fuel Building ML20134A9571997-01-16016 January 1997 Proposed Tech Specs LCO 3.1.4 Re Spent Fuel Pool Load Restrictions ML20135B5071996-11-27027 November 1996 Proposed Tech Specs Reflecting New Fuel Debris Storage Container Design & Storage of non-fuel Bearing Components & Fuel Rod Storage Container in Trojan ISFSI ML20129E7341996-10-23023 October 1996 Proposed Tech Specs Re Spent Fuel Pool Debris Processing ML20059L1721993-11-30030 November 1993 Final Update to PGE-1043,Amend 3, Accident Monitoring Instrumentation Review, Identifying Rept as Historical by Placing Explanation in Preface to Be Inserted at Front of Document ML20059A5601993-10-21021 October 1993 Proposed Tech Specs Modifying Security License Conditions for Clarity ML20057D9131993-09-30030 September 1993 Amend 14 to PGE-1012, Trojan Nuclear Plant Fire Protection Plan. W/Four Oversize Drawings ML20057F1631993-09-30030 September 1993 Updated Trojan Nuclear Div Manual ML20056F9971993-08-26026 August 1993 Rev 9 to Odcm ML20127P2121993-01-27027 January 1993 Proposed TS 6.2.2.f Re Composition of Fire Brigade & TS Table 6.2-1 Re Min Shift Crew Composition for Shift Manager & Noncertified Operator ML20126B7231992-12-0303 December 1992 Rev 13 to Chemistry Manual Procedure CMP 1, Primary-to-Secondary Leak Rate ML20127M6391992-11-16016 November 1992 Proposed Tech Specs for Deferral of Unscheduled Steam Generator ISI ML20126B7601992-09-25025 September 1992 Rev 18 to Emergency Instruction EI-3, SG Tube Rupture ML20126B7451992-09-25025 September 1992 Rev 7 to Event-Specific Emergency Instruction ES-3.1, Post- SG Tube Rupture Cooldown Using Backfill ML20126B7491992-09-25025 September 1992 Rev 7 to Event-Specific Emergency Instruction ES-3.3, Post- SG Tube Rupture Cooldown Using Steam Dump ML20126B7411992-09-25025 September 1992 Rev 8 to Event-Specific Emergency Instruction ES-3.2, Post- Steam Generator Tube Rupture Cooldown Using SG Blowdown ML20126B7331992-09-25025 September 1992 Rev 7 to Emergency Contingency Action Procedure ECA-3.3, SG Tube Rupture W/O Pressurizer Pressure Control ML20126B7061992-09-25025 September 1992 Rev 7 to Emergency Contingency Action Procedure ECA-3.2, SG Tube Rupture W/Loss of Reactor Coolant - Saturated Recovery Desired ML20126B7181992-09-23023 September 1992 Rev 7 to Emergency Contingency Action Procedure ECA-3.1, SG Tube Rupture W/Loss of Reactor Coolant - Subcooled Recovery Desired ML20126B7141992-08-20020 August 1992 Rev 19 to Off-Normal Instruction ONI-12, High Activity Radiation Monitoring ML20210D5281992-06-0808 June 1992 Errata to Amend 6 to PGE-1025, Environ Qualification Program Manual ML20141M1901992-03-27027 March 1992 Proposed Tech Specs Re Surveillance Requirements for Lco. Response to Request for Addl Info on Subj Encl ML20126B7201992-01-24024 January 1992 Rev 2 to Off-Normal Instruction ONI-3-12, SG Tube Leak ML20062F7391990-11-15015 November 1990 Amend 4 to PGE-1049, Trojan Nuclear Plant Inservice Insp Program for Second 10-Yr Interval,Jan 1987 - Dec 1996 ML20059L6701990-09-21021 September 1990 Proposed Tech Spec Table 3.6-1, Containment Isolation Valves ML20247C8941989-07-19019 July 1989 Proposed Tech Spec Table 3.6-1, Containment Isolation Valves ML20245C8091989-06-12012 June 1989 Proposed Tech Specs,Changing Orifice Size of Steam Line Safety Valves Shown in Tech Spec Table 4.7-1, Steam Line Safety Valves Per Loop ML20247E5331989-03-29029 March 1989 Rev 0 to Procedure QTS-VT-100, Visual Exam Procedure ML20248D3271989-03-27027 March 1989 Rev 0 to QAP-VT-108, Visual Exam ML20247E5491989-03-0202 March 1989 Rev 0 to Procedure TNP-QAP-VT108, Visual Exam ML20235Q9191989-02-10010 February 1989 Proposed Tech Specs Re Limiting Conditions for Operation & Surveillance Requirements for Containment Integrity for Equipment Hatches ML20235L9441989-02-10010 February 1989 Proposed Tech Specs,Allowing Mods to Control Auxiliary Fuel Bldg Complex Resulting in Increase in Lateral Shear Forces on Any Story ML20154D1781988-09-0909 September 1988 Proposed Tech Specs Revising Pressurizer LPSI Setpoint 1999-08-05
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o
. , LCA 1130 Rev. 1 o Attachment 1 Page 1 of 5 s
INDEX BASES SECTION Page 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION. . . . . . . . . . . . . . B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION . . . . . . B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION. . . . . . . . . . . . . . B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS . . . . . . . . . . . . . . . . B 3/4 4-1 3/4.4.2 and 3/4.4 3 SAFETY AND RELIEF VALVES . . . . . . . . . B 3/4 4-la l 3/4.4.4 PRESSURIZER . . . . . . . . . . . . . . . . . . . . . B 3/4 4-2 j 3/4.4.5 STEAM GENERATORS. . .;. . . . . . .......... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE. . . . . . . . . . . . B 3/4 4-2 3/4.4.7 CHEMISTRY . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.8 SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.9 PRESSURE / TEMPERATURE LIMITS . . . . . . . . . . . . . B 3/4 4-4 3/4.4.10 STRUCTURAL INTEGRITY. ................ B 3/4 4-10 3/4.4.11 REACTOR VESSEL HEAD VENTS . . . . . . . . . . . . . . B 3/4 4-10 7
a 8512310352 851227 PDR ADOCK 05000344 P PDR TROJAN-UNIT 1 X Amendment No. 93,
a
. , LCA 113, Rev. 1 Attachment 1 Page 2 of 5 REACTOR COOLANT SYSTEM ACTION (Continued) b) Place the following reautor trip system and ESFAS instrumentation channels, associated with the loop not in operation, in their tripped conditions:
- 1) Overpower AT channel.
- 2) Overtemperature AT channel.
- 3) - Low-Low channel used in the coincidence Tavg circu it with steam Flow - High for Safety Injection.
- 4) Steam Line Pressure - Low channel used in the coincidence circuit with Steam Flow - High for Safety Injection.
- 5) Steam Flow-High channel used for Safety Injection.
- 6) Differential Pressure Between Steam Lines - High Channel used for Safety Injection (trip all l bistables which indicate low active loop steam pressure with respect to the idle loop steam pressure).
c) Change the P-8 interlock setpoint from the value specified in Table 3.3-1 to 175% of RATED THERMAL POWER.
- 2. THERMAL POWER is restricted to $70% of RATED THERMAL POWER.
Below P-7:
- a. With Keff 11.0, operation below P-7 may proceed provided'at least four reactor coolant loops and associated pumps are in l operation.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
TROJAN-UNIT 1 3/4 4-2 Amendment No. 54 i
i
, Attachment 1 i Page 3 of 5 '
REACTOR COOLANT SYSTEM l HOT STANDBY
)
LIMITING CONDITION FOR OPERATION 3.4.1.2 a. With any control rod drive mechanism (CRDM) energized, all four reactor coolant loops shall be in operation.
- b. With no CRDMs energized, at least two reactor coolant loops shall be OPERABLE with one reactor coolant loop in operation.*
APPLICABILITY: MODE 3 ACTION:
a.
With less than de-energize four reactor coolant loops in operation, imediately {
all CRDMs.
- b. With less than two reactor coolant loops OPERABLE when the CRDMs are de-energized, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c. With no reactor coolant 1 cop in operation, suspend all operations [
involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
SURVEILLANCE REQUIREMENTS 4 . 4 .1. 2.1 All CRDMs shall be verified de-energued at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if less than four reactor coolant loops are in operation.
4.4.1.2.2 At least two reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker [
alignments and indicated power availability.
4.4.1.2.3 At least one cooling loop shall be verified to be in operation I and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is
- maintained at 10*F below saturation temperature.
l TROJAN-UNIT 1 3/4 4-2b Amendment No. H
LCA 113, Rev. 1
. Attachment 1 Page 4 of 5 3/4.4 REACTOR COOLANT SYSTEM BASES i 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.73 during all normal operations and anticipated transients. With one reactor coolant loop not in operation, THERMAL POWER is restricted to <58 percent of RATED THERMAL POWER until the Overtemperature AT trip is reset. Either action ensures that the DNBR will be maintained above 1.73. A loss of flow in two loops will cause a reactor trip if operating above F-7 (10 percent of RATED THERf1AL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (39 percent of RATED THERMAL POWER).
In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removitig decay heat; however, single failure con-siderations require that two loops be OPERABLE. Four loops in operation while control rod drive mechanisms are energized ensures that the DNB design basis can be met for a bank withdrawal from subcritical or low power accident.
In MODES 4 and 5, a single reactor coolant loop or RHR loop pro-vides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR pump provides adaquate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 290*F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water vol-ume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures, or (3) by restricting starting of an RCP unless another RCP is running.
TROJAN-UNIT 1 B 3/4 4-1 Amendment No. 5#, 55, 78
/ .
LCA 113, Rsv. 1 o Attachment 1 Page 5 of 5 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES and 3/4.4.3 SAFETY AND RELIEF VALVES l The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 pSig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at 110% of the valve's setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, pror! des overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (ie, no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power-operated relief valves or steam dump valves.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the p$ovisions of Section XI of the ASME Boiler and Pressure Code.
TROJAN-UNIT 1 B 3/4 4-la Amendment No. 5#, 7#
r UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )
)
PORTLAND CENERAL ELECTRIC COMPANY, ) Docket 50-344 THE CITY OF EUGENE, OREGON, AND ) Operating License WPF-1 PACIFIC POWER & LICHT COMPANY )
)
(TROJAN NLCLEAR PLANT) )
CERTIFICATE OF SERVICE I hereby certify that copies of License Change Application 113 Revision 1 to the Operating License for Trojan Nuclear Plant, dated December 27, 1985, have been served on the following by hand delivery or by deposit in the United States mail, first class, this 27th day of December 1985:
Mr. Lynn Frank, Director State of Oregon Department of Energy Labor & Industries Blds, Rm 102 Salem OR 97310 Mr. Robert L. King Chairman of County Commissioners Columbia County Courthouse St. Helens OR 97051
/v .w A. Zinmerman, Manager lear Regulation Branch ear Safety & Regulation Subscribed and sworn to before me this 27th day of December 1985.
jh )
NOTA 2 PUGU$ EGON t.ty commis: ion txpes. N'8f88
,g,g w____-.