Similar Documents at Surry |
---|
Category:TECHNICAL SPECIFICATIONS
MONTHYEARML18152B3591999-08-23023 August 1999 Revised Tech Specs Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in UFSAR & Includes Ref to Applicable UFSAR Section ML18152B6591999-04-28028 April 1999 Proposed Tech Specs Re Refueling Water Chemical Addition Tank Min Vol ML18152A2131999-02-16016 February 1999 Proposed Tech Specs,Consolidating AFW cross-connect Requirements by Relocation of Electrical Power Requirements from TS 3.16 to TS 3.6 ML18152B5451999-02-16016 February 1999 Proposed Tech Specs Pages,Revising Augmented Insp Requirements for Reactor Coolant Pump Flywheels ML18152B6151998-11-0404 November 1998 Proposed Tech Specs 4.6.A.1.b,re EDG Start & Load Time Testing Requirements & TS 3.16 Bases Re EDG Ratings ML18153A3311998-09-24024 September 1998 Proposed Tech Specs Modifying Testing Requirements for Reactor Trip Bypass Breaker ML18153A3351998-09-24024 September 1998 Proposed Tech Specs Pages Affected by Suppl to 960912 Resubmittal of Change Request Re Relocation of Fire Protection Requirements from TS to UFSAR ML18152B7551998-06-19019 June 1998 Proposed Tech Specs Establishing Requirements for Use of Temporary Supply Line (Jumper) to Provide Svc Water to Component Cooling Heat Exchangers ML18152A3651998-03-25025 March 1998 Proposed Tech Specs Revising Station Mgt Titles to Reflect New Positions Approved by Vepc Board of Directors on 980220 ML18153A3481997-12-18018 December 1997 Proposed Tech Specs Clarifying Terminology Used for Describing Equipment Surveillances Conducted on Refueling Interval Frequency.Clarification Consistent W/Info Contained in Rev 1 to NUREG-1431 ML18153A1761997-11-0505 November 1997 Proposed Tech Specs Re Temporary Svc Water Supply Line to Component Cooling Heat Exchangers ML18153A3941997-11-0505 November 1997 Proposed Tech Specs Re Change for Increased Enrichment of Reload Fuel ML18153A5231997-04-24024 April 1997 Proposed Corrected Tech Specs Pages 6.1-3 & 6.1-8 Re Relocation of Fire Protection TS to Updated Final Safety Analysis Rept ML18153A5031997-03-18018 March 1997 Proposed Tech Specs Rev to Section 4.15 for Surry Power Station to Include Pp Inadvertently Omitted from 970203 Request for Amend to Licenses DPR-32 & DPR-37 ML18153A4921997-02-0303 February 1997 Proposed Tech Specs Re Deletion of Specific ASME Section XI Code Ref ML18153A6351996-11-26026 November 1996 Proposed Tech Specs Re Removal of Record Retention Requirements,Per GL 95-06 & Administrative Ltr 95-06 ML18153A0671996-09-12012 September 1996 Proposed Tech Specs Re Relocation of Fire Protection Requirements ML18153A6901996-04-15015 April 1996 Proposed Tech Specs,Clarifying Applicability of Quadrant Power Tilt Ration Requirements ML18153A5391996-03-21021 March 1996 Proposed Tech Specs Re Charcoal Filter Testing Clarification ML18153A5271996-03-14014 March 1996 Proposed Tech Specs,Permitting Use of 10CFR50 App J,Option B,performance-based Containment Lrt ML18153A5801996-01-30030 January 1996 Proposed Tech Specs Re Reactor Coolant Sys Liquid Sampling ML18153A6761995-11-20020 November 1995 Proposed Tech Specs Re App J Option B,performance-based Containment Leakage Rate Testing ML18153A7141995-07-20020 July 1995 Proposed Tech Specs Establishing New Setpoint Limit for SG high-high Level & Provides More Restrictive Setting Limits for Certain Rps/Esfas Setpoints ML18153A6991995-07-14014 July 1995 Proposed Tech Specs,Providing Two H Allowed Outage Time for One RHR Pump to Accommodate Plant Safety,Emergency Power Sys Surveillance Testing & Permit Depressurizing SI Accumulators in Lieu of Accumulator Isolation ML18153A8371995-06-0808 June 1995 Proposed Tech Specs,Incorporating Revised Pressure/Temp Limits & Associated Ltops Setpoint That Will Be Valid to end-of-license ML18153B2301995-02-14014 February 1995 Proposed Tech Specs Re App J Testing Requirements ML18153B2131995-01-24024 January 1995 Proposed Tech Specs,Modifying as-found Test Acceptance Criterion for Pressurizer Safety Valves ML18153B1621994-11-29029 November 1994 Proposed Tech Specs Implementing Zirlo Fuel Cladding ML18153B1581994-11-22022 November 1994 Proposed Tech Specs,Deleting Unnecessary Descriptive Phrases Re Number of Cells in Station & EDG Batteries ML18153B1501994-11-10010 November 1994 Proposed Tech Specs Re Changes to TS Will Clarify SR for Reactor Protection & Engineered Safeguard Sys Instrumentation & Actuation Logic ML18153B0941994-10-11011 October 1994 Proposed Tech Specs Surveillance Frequencies for Hydrogen Analyzers ML18152A5061994-09-0606 September 1994 Proposed Tech Specs Re Mgt Safety Review Committee & Station Nuclear Safety & Operating Committee Responsibilities ML18152A1191994-08-30030 August 1994 Proposed Tech Specs to Accomodate Core Uprating ML18153B0061994-07-14014 July 1994 Proposed Tech Specs Changes Will Eliminate Remaining References to cycle-specific Parameters in Surry TS ML18152A1821994-06-0909 June 1994 Proposed TS Re Chemical & Vol Control Sys & Safety Injection Sys ML20065P6111994-04-19019 April 1994 Proposed Tech Specs,Modifying Control Rod Movement Surveillance Frequency ML18153A8821994-02-25025 February 1994 Proposed Tech Specs Modifying Surveillance Frequency of Nozzles in Containment & Recirculation Spray Sys ML18153B4321993-12-27027 December 1993 Proposed Tech Specs Reflecting Revised Review Responsibilities of Station Nuclear Safety & Operating Committee & Mgt Safety Review Committee ML18153B4081993-12-10010 December 1993 Proposed Tech Specs,Establishing Upper Limit on Allowable COLR Mtc Values ML18153B4021993-12-10010 December 1993 Proposed Tech Specs,Modifying Surveillance Frequency of Auxiliary Feedwater Sys Pumps from Monthly to Quarterly Per Generic Ltr 93-05 & NUREG-1431 ML18153B4051993-12-10010 December 1993 Proposed Tech Specs,Updating Augmented Insp Program for Sensitized Stainless Steel by Incorporating Newer Code Requirements,While Maintaining Augmented Insp Philosophy ML18153B3751993-11-15015 November 1993 Proposed Tech Specs Implementing Revised 10CFR20,revising Frequency Radiological Release Repts from Semiannual to Annual & Clarifying Site Maps ML18153B3501993-10-19019 October 1993 Proposed Tech Specs for RSHX Svc Water Outlet Radiation Monitors ML18153B3321993-09-29029 September 1993 Proposed Tech Specs Modifying Required Insp Frequency of Low Pressure Turbine Blades to Permit Blade Insp to Be Performed Concurrent W/Disk & Hub Insp ML18152A0481993-07-20020 July 1993 Proposed Tech Specs Deleting Requirement for Station Nuclear Safety & Operating Committee & Audit Frequencies ML18153D3921993-07-16016 July 1993 Proposed Tech Specs for Operation W/Three Degree Increase in Svc Water Temp Limit for Containment Air Partial Pressures of 9.1,9.2 & 9.35 Psia ML18152A0461993-07-16016 July 1993 Proposed Tech Specs Implementing Revised 10CFR20,revise Frequency of Radiological Effluent Release Repts from semi-annual to Annual,& Clarify Site Maps ML18152A4511993-07-0202 July 1993 Proposed Tech Specs to Include COLR Which Presents reload- Specific Limits for Key Core Operating Parameters ML18153D3811993-07-0202 July 1993 Proposed TS Table 4.2-1 Re Miscellaneous Insps & Sensitized Stainless Steel Exams ML18153D3331993-05-0606 May 1993 Proposed Tech Specs Supporting Operation of Unit 2 w/100 Psi Reduction in RCS Nominal Operating Pressure Through End of Operating Cycle 12 1999-08-23
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML18152B3591999-08-23023 August 1999 Revised Tech Specs Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in UFSAR & Includes Ref to Applicable UFSAR Section ML18151A6641999-08-0606 August 1999 Rev 0 to Surry Unit 2 Cycle 16 Startup Physics Tests Rept. ML18152B6591999-04-28028 April 1999 Proposed Tech Specs Re Refueling Water Chemical Addition Tank Min Vol ML18152A2131999-02-16016 February 1999 Proposed Tech Specs,Consolidating AFW cross-connect Requirements by Relocation of Electrical Power Requirements from TS 3.16 to TS 3.6 ML18152B5451999-02-16016 February 1999 Proposed Tech Specs Pages,Revising Augmented Insp Requirements for Reactor Coolant Pump Flywheels ML18151A5511999-02-10010 February 1999 to NE-1187, Surry Unit 1,Cycle 16 Startup Physics Tests Rept. ML18152B6151998-11-0404 November 1998 Proposed Tech Specs 4.6.A.1.b,re EDG Start & Load Time Testing Requirements & TS 3.16 Bases Re EDG Ratings ML18153A3351998-09-24024 September 1998 Proposed Tech Specs Pages Affected by Suppl to 960912 Resubmittal of Change Request Re Relocation of Fire Protection Requirements from TS to UFSAR ML18153A3311998-09-24024 September 1998 Proposed Tech Specs Modifying Testing Requirements for Reactor Trip Bypass Breaker ML18152B7551998-06-19019 June 1998 Proposed Tech Specs Establishing Requirements for Use of Temporary Supply Line (Jumper) to Provide Svc Water to Component Cooling Heat Exchangers ML20249B9911998-05-0606 May 1998 Analysis of Capsule X Virginia Power Surry Unit 1 Reactor Vessel Matl Surveillance Program. W/Evaluation of Surry Unit 1 Surveillance Capsule X Results & Response to NRC RAI Re GL 92-01,rev 1,suppl 1 ML18151A1931998-05-0404 May 1998 Rev 1 to Summary of Changes to Surry Units 1 & 2 Third Interval IST Program. ML18152A3651998-03-25025 March 1998 Proposed Tech Specs Revising Station Mgt Titles to Reflect New Positions Approved by Vepc Board of Directors on 980220 ML20199B0711998-01-0505 January 1998 Rev 0 to NE-1148, Surry Unit 2,Cycle 15 Startup Physics Test Rept ML18153A3481997-12-18018 December 1997 Proposed Tech Specs Clarifying Terminology Used for Describing Equipment Surveillances Conducted on Refueling Interval Frequency.Clarification Consistent W/Info Contained in Rev 1 to NUREG-1431 ML18150A4661997-12-16016 December 1997 ISI Plan for Third Insp Interval,Vol 2,Rev 9 for Components & Component Supports,940510-040510, for Surry Power Station,Unit 2 ML18153A3941997-11-0505 November 1997 Proposed Tech Specs Re Change for Increased Enrichment of Reload Fuel ML18153A1761997-11-0505 November 1997 Proposed Tech Specs Re Temporary Svc Water Supply Line to Component Cooling Heat Exchangers ML18150A4641997-10-27027 October 1997 Risk-Informed ISI (RI-ISI) Pilot Program Submittal. ML18151A3911997-10-16016 October 1997 Rev 8 to VPAP-2103, Odcm. ML18151A7231997-08-0707 August 1997 Rev 1 to Nuclear Safety Analysis Manual Part Iv,Chapter a Probabilistic Safety Assessment Products. ML20210J5031997-07-31031 July 1997 Rev 0 to NE-1132, Surry Unit 1,Cycle 15 Startup Physics Tests Rept ML18150A4441997-06-0909 June 1997 Vol 2,Rev 8 to ISI Plan for Third Insp Interval for Components & Component Supports,Oct 14,1993-Oct 13,2003. ML18153A5231997-04-24024 April 1997 Proposed Corrected Tech Specs Pages 6.1-3 & 6.1-8 Re Relocation of Fire Protection TS to Updated Final Safety Analysis Rept ML18153A5031997-03-18018 March 1997 Proposed Tech Specs Rev to Section 4.15 for Surry Power Station to Include Pp Inadvertently Omitted from 970203 Request for Amend to Licenses DPR-32 & DPR-37 ML18153A4921997-02-0303 February 1997 Proposed Tech Specs Re Deletion of Specific ASME Section XI Code Ref ML18153A6351996-11-26026 November 1996 Proposed Tech Specs Re Removal of Record Retention Requirements,Per GL 95-06 & Administrative Ltr 95-06 ML18153A0671996-09-12012 September 1996 Proposed Tech Specs Re Relocation of Fire Protection Requirements ML18151A9761996-08-13013 August 1996 Cycle 14 Startup Physics Test Rept. W/960830 Ltr ML20134J9861996-07-30030 July 1996 /Unit 2 Fuel Assembly Insp Program ML18152A4701996-06-13013 June 1996 Cycle 13 Control Rod Performance Test Results. ML18153A6901996-04-15015 April 1996 Proposed Tech Specs,Clarifying Applicability of Quadrant Power Tilt Ration Requirements ML18153A5391996-03-21021 March 1996 Proposed Tech Specs Re Charcoal Filter Testing Clarification ML18153A5271996-03-14014 March 1996 Proposed Tech Specs,Permitting Use of 10CFR50 App J,Option B,performance-based Containment Lrt ML18153A5801996-01-30030 January 1996 Proposed Tech Specs Re Reactor Coolant Sys Liquid Sampling ML18152A0571995-12-20020 December 1995 Startup Physics Test Rept,Surry Unit 1,Cycle 14. W/960111 Ltr ML18153A6761995-11-20020 November 1995 Proposed Tech Specs Re App J Option B,performance-based Containment Leakage Rate Testing ML18151A6421995-08-0101 August 1995 Change 3 to Rev 0 to Third Interval IST Program ML18153A7141995-07-20020 July 1995 Proposed Tech Specs Establishing New Setpoint Limit for SG high-high Level & Provides More Restrictive Setting Limits for Certain Rps/Esfas Setpoints ML18153A6991995-07-14014 July 1995 Proposed Tech Specs,Providing Two H Allowed Outage Time for One RHR Pump to Accommodate Plant Safety,Emergency Power Sys Surveillance Testing & Permit Depressurizing SI Accumulators in Lieu of Accumulator Isolation ML18153A8371995-06-0808 June 1995 Proposed Tech Specs,Incorporating Revised Pressure/Temp Limits & Associated Ltops Setpoint That Will Be Valid to end-of-license ML20083C9951995-05-0808 May 1995 Rev 0 to Surry Unit 2,Cycle 13 Startup Physics Tests Rept ML18153B2301995-02-14014 February 1995 Proposed Tech Specs Re App J Testing Requirements ML18153B2131995-01-24024 January 1995 Proposed Tech Specs,Modifying as-found Test Acceptance Criterion for Pressurizer Safety Valves ML18153B1621994-11-29029 November 1994 Proposed Tech Specs Implementing Zirlo Fuel Cladding ML18153B1581994-11-22022 November 1994 Proposed Tech Specs,Deleting Unnecessary Descriptive Phrases Re Number of Cells in Station & EDG Batteries ML18153B1501994-11-10010 November 1994 Proposed Tech Specs Re Changes to TS Will Clarify SR for Reactor Protection & Engineered Safeguard Sys Instrumentation & Actuation Logic ML18153B0941994-10-11011 October 1994 Proposed Tech Specs Surveillance Frequencies for Hydrogen Analyzers ML18152A5061994-09-0606 September 1994 Proposed Tech Specs Re Mgt Safety Review Committee & Station Nuclear Safety & Operating Committee Responsibilities ML18152A1191994-08-30030 August 1994 Proposed Tech Specs to Accomodate Core Uprating 1999-08-06
[Table view] |
Text
- TS 2.3-3 where 8 T O = Indicated 8 T at rated thermal power, °F T = Average coolant temperature, °F T' = Average coolant temperature measured at nominal conditions and rated power, °F K4 = A constant= 1.089 K5 = O for decreasing average temperature A constant, for increasing average temperature 0.02/°F K6 = 0 for T:s:;T'
= 0.001086 for T > T' f(81) as defined in (d) above,
't" = 10 seconds 3
(f) Low reactor coolant loop flow = ~ 90% of normal indicated loop flow as measured at elbow taps in each loop (g) Low reactor coolant pump motor frequency - ~ 57.5 Hz (h) Reactor coolant pump under voltage - ~ 70% of normal voltage
- 3. Other reactor trip settings (a) High pressurizer water level - ::;; 92% of span (b) Low-low steam generator water level - ~ 14.5% of narrow range instrument span (c) Low steam generator water level - ~ 15% of narrow range instrument span in coincidence with steam/feedwater mismatch flow - ::;; 1.0 x 106 lbs/hr (d) Turbine trip (e) Safety injection - Trip settings for Safety Injection are detailed in TS Section 3.7.
Amendment Nos.
9507270070 950720 PDR ADOCK 05000280 I '
P PDR
- TS 2.3-4 B. Protective instrumentation settings for reactor trip interlocks shall be as follows:
- 1. The reactor trip on low pressurizer pressure, high pressurizer level, turbine trip, and low reactor coolant flow for two or more loops shall be unblocked when power ~ 10% of rated power.
- 2. The single loop loss of flow reactor trip shall be unblocked when the power range nuclear flux~ 50% of rated power.
- 3. The power range high flux, low setpoint trip and the intermediate range high flux, high setpoint trip shall be unblocked when power
- 10% of rated power.
- 4. The source range high flux, high setpoint trip shall be unblocked when the intermediate range nuclear flux is ::;; 5 x 10-11 amperes.
The power range reactor trip low setpoint provides protection in the power range for a power excursion beginning from low power. This trip value was used in the safety analysis. ( 1) The Source Range High Flux Trip provides reactor core protection during shutdown (COLD SHUTDOWN, INTERMEDIATE SHUTDOWN, and HOT SHUTDOWN) when the reactor trip breakers are closed and reactor power is below the permissive P-6. The Source and Intermediate Range trips in addition to the Power Range trips provide core protection during Amendment Nos.
e
- TS 2.3-8 will prevent the minimum value of the DNBR from going below the applicable design as a result of the decrease in Reactor Coolant System flow associated with the loss of a single reactor coolant pump.
Although not necessary for core protection, other reactor trips provide additional protection. The steam/feedwater flow mismatch which is coincident with a low steam generator water level is designed for and provides protection from a sudden loss of the reactor's heat sink. Upon the actuation of the safety injection circuitry, the reactor is tripped to decrease the severity of the accident condition.
Upon turbine trip, at greater than 10% power, the reactor is tripped to reduce the severity of the ensuing transient.
References (1) FSAR Section 14.2.1 (2) FSAR Section 14.2 (3) FSAR Section 14.5 (4) FSAR Section 7 .2 (5) FSAR Section 3.2.2 (6) FSAR Section 14.2.9 (7) FSAR Section 7 .2 Amendment Nos.
.'
-
- TS 3.7-6 reduces the consequences of a steam line break inside the containment by stopping the entry of feedwater.
Auxiliary Feedwater System Actuation The automatic initiation of auxiliary feedwater flow to the steam generators by instruments identified in Table 3.7-2 ensures that the Reactor Coolant System decay heat can be removed following loss of main feedwater flow. This is consistent with the requirements of the "TMl-2 Lessons Learned Task Force Status Report," NUREG-0578, item 2.1.7.b.
Setting Limits
- 1. The high containment pressure limit is set at about 10% of design containment pressure. Initiation of safety injection protects against loss of coolant(2) or steam line break(3) accidents as discussed in the safety analysis.
- 2. The high-high containment pressure limit is set at about 23% of design containment pressure. Initiation of containment spray and steam line isolation protects against large loss-of-coolant(2) or steam line break accidents(3) as discussed in the safety analysis.
- 3. The pressurizer low pressure setpoint for safety injection actuation is set substantially below system operating pressure limits. However, it is sufficiently high to protect against a loss-of-coolant accident as shown in the safety analysis. (2) The setting limit (in units of psig) is based on nominal atmospheric pressure.
- 4. The steam line high differential pressure limit is set well below the differential pressure expected in the event of a large steam line break accident as shown in the safety analysis. (3)
- 5. The high steam line flow differential pressure setpoint is constant at 40%
full flow between no load and 20% load and increasing linearly to 110%
of full flow at full load in order to protect against large steam line break accidents. The coincident low T avg setting limit for SIS and steam line isolation initiation is set below its HOT SHUTDOWN value. The coincident Amendment Nos.
TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING No. Functional Unit Channel Action Setting Limit 1 High Containment Pressure (High Containment a) Safety Injection ~ 19 psia Pressure Signal) b) Containment Vacuum Pump Trip
,*
c) High Press. Containment Isolation d) Safety Injection Containment Isolation e) F.W. Line Isolation 2 High-High Containment Pressure (High-High a) Containment Spray ~25 psia Containment Pressure Signals) b) Recirculation Spray c) Steam Line Isolation d) High-High Press. Containment Isolation 3 Pressurizer Low-Low Pressure a) Safety Injection ~ 1,760 psig b) Safety Injection Containment Isolation c) F.W. Line Isolation 4 High Differential Pressure Between a) Safety Injection ~ 150 psig Steam Line and the Steam Line Header b) Safety Injection Containment Isolation c) F.W. Line Isolation
- i::,, ~ 40% (at zero load) of full 3 5 High Steam Flow in 2/3 Steam Lines a) Safety Injection steam flow
- ,
c.. ~ 40% (at 20% load) of full 3
CD steam flow
- ,
C"+ ~ 110% (at full load) of full 2 steam flow
.
0 Cll b) Steam Line Isolation c) Safety Injection Containment Isolation d) F.W. Line Isolation Coincident with Low Tavg or ~541°FTavg Low Steam Line Pressure ~ 500 psig steam line pressure -I V,
.w
'-I I
N c.,,
TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING No. Functional Unit Channel Action Setting Limit 6 AUXILIARY FEEDWATER
- a. Steam Generator Water Level Low-Low Aux. Feedwater Initiation 2:'. 14.5% narrow range SIG Slowdown Isolation
- b. RCP Undervoltage Aux. Feedwater Initiation 2:'. 70% nominal C.
d.
e.
Safety Injection Station Blackout Main Feedwater Pump Trip Aux. Feedwater Initiation Aux. Feedwater Initiation Aux. Feedwater Initiation All S.I. setpoints
~
N.A.
46.7% nominal **
7 LOSS OF POWER
- a. 4.16 KV Emergency Bus Undervoltage Emergency Bus Separation and 75 {+/- 1.0)% volts with a (Loss of Voltage) Diesel start 2 (+5, -0.1) second time delay
- b. 4.16 KV Emergency Bus Undervoltage Emergency Bus Separation and 90 (+/- 1)% volts with a (Degraded Voltage) Diesel start 60 (+/- 3.0) second time delay (Non CLS, Non SI)
> 7 (+/- .35) second time delay a (CLS or SI Conditions)
- ,
Q.
a Cl) 8 NON-ESSENTIAL SERVICE WATER ISOLATION
- ,
c-+-
- a. Low Intake Canal Level Isolation of Service Water flow to 23 feet-6 inches
- z non-essential loads
.
0 1/)
9 RECIRCULATION MODE TRANSFER
- a. RWST Level-Low Initiation of Recirculation Mode 2:'. 18.93%
Transfer System ~19.43%
-I V'I 10 TURBINE TRIP AND FEEDWATER ISOLATION
- a. Steam Generator Water Level High-High Turbine Trip ~ 80% narrow range
.
<.,.)
'-I I
Feedwater Isolation N C'\
-
- ATTACHMENT 3 SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
-~
10
~ .
CFR 50.92 EVALUATION - BASIS FOR NO SIGNIFICANT HAZARDS
- DETERMINATION Virginia Electric and Power Company is proposing revisions to Sections 2.3, "Limiting Safety System Settings, Protective Instrumentation," and 3.7, "Instrumentation Systems." We have reviewed the proposed change against the criteria of 10 CFR 50.92 and have concluded that the change does not pose a significant safety hazards consideration as defined therein. Consistent with the examples of amendments not likely to involve a significant hazards consideration noted in the Federal Register (Vol. 50, No. 44) dated March 6, 1986, the proposed changes to 1) revise the units of the high-high containment pressure setpoint limit and 2) delete certain references to two-loop operation, since the plant is not licensed to operate in that manner, are purely administrative in nature and therefore are not a significant hazards consideration.
Likewise, the remaining proposed changes constitute additional restrictions not presently included in Technical Specifications and therefore are not a significant hazards consideration. Specifically, operation of Surry Power Station with the proposed change will not:
- 1. Involve a significant increase in either the probability of occurrence or consequences of any accident or equipment malfunction scenario which is important to safety and which has been previously evaluated in the Updated Safety Analysis Report (UFSAR). The effect of the proposed change is to ensure that actual plant setpoints remain conservative consistent with respect to accident analysis assumptions. The proposed change requires safety system actuation limits that are more conservative than those currently in Technical Specifications. The change does not invalidate currently implemented station setpoints or currently applicable accident analysis assumptions regarding these setpoints. Consequently, the results and conclusions of the current UFSAR accident analyses are not affected by these changes. The proposed Technical Specifications change revises setpoints used to mitigate accidents and therefore has no bearing on the probability of an accident. Further, the change ensures that the setpoints used to mitigate an accident bound the setpoints used in the accident analyses. Therefore, the probability of an accident or consequences of an accident is not adversely affected as a result of this change.
- 2. Create the possibility of a new or different type of accident than those previously evaluated in the UFSAR. Implementing the proposed Technical Specifications setpoint limits cannot create the possibility of an accident of a different type than was previously evaluated in the UFSAR. Since actual plant setpoints are not being affected, new accident precursors will not be introduced. Furthermore, spurious challenges to safety systems are also not expected to increase in frequency as a result of these changes since actual setpoints installed in the plant are not being changed. Consequently, no new accident precursors are created as a result of the new Technical Specifications setpoint limits.
3.,
. .
~
-
- Involve a significant reduction in a margin of safety. Since the results of the existing UFSAR accident analyses remain bounding, safety margins are not impacted. The proposed Technical Specifications setpoint limits ensure plant setpoints remain conservative and consistent with design base accident analysis assumptions including appropriate instrument channel uncertainties due to harsh environmental conditions. Therefore, the margin of safety as defined in the Technical Specifications bases is unaffected.