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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML17284A9121999-10-13013 October 1999 Proposed Tech Specs 3.3.6.1,Table 3.3. 6.1-1, Primary Containment Isolation Instrumentation. ML17284A8521999-07-29029 July 1999 Proposed Tech Specs 3.4.9, RHR Shutdown Cooling Sys - Hot Shutdown. ML17284A8491999-07-29029 July 1999 Proposed Tech Specs,Revising SR 3.5.2.2 Re Condensate Storage Tank Water Level ML17284A8461999-07-29029 July 1999 Proposed Tech Specs,Revising Table 3.3.5.1-1, ECCS Instrumentation Items 1.a,2.a,4.a & E.A. ML17284A8421999-07-29029 July 1999 Proposed Tech Specs Revising SR of TS 3.8.4, DC Sources - Operating & SR 3.8.5.1 of TS 3.8.5, DC Sources - Shutdown. ML17333A0021999-04-20020 April 1999 Proposed Tech Specs Section 3.4.11,replacing Existing Reactor Pressure Temp Limit Curves by 000630 ML17292B6341999-04-0707 April 1999 Proposed Tech Specs Modifying MCPR Safety Limits to Allow Continued Power Operation at Plant Following Restart from R-14 RFO ML17292B4881998-12-17017 December 1998 Proposed Tech Specs SR 3.8.1.8,allowing Capability to Manually Transfer Between Preferred & Alternate Offsite Power Sources During Modes 1 & 2 by 990125 ML20198A7051998-11-30030 November 1998 Revs 8 Through 13 to TS Bases & Revs 12 Through 15 of Licensee Controlled Specs ML17284A7181998-08-0505 August 1998 Proposed Tech Specs SR 3.8.4.7,allowing Performance Discharge Test to Be Performed in Lieu of Battery Svc Test for Div 1 & 3,125 Vdc Batteries E-B1-1 & HPCS-B1-DG3 & Div 1,250 Vdc Battery E-B2-1 ML17284A7071998-07-17017 July 1998 Proposed Tech Specs Modifying SR 3.8.4.7 to Allow Performance Discharge Test to Be Performed in Lieu of Battery Svc Test for Div 2,125 Vdc,Battery E-B1-2 ML17292B2831998-03-0909 March 1998 Proposed Tech Specs Establishing Interim SLMCPR for Siemens Power Corp ATRIUM-9X Fuel Applicable to Cycle 14 Only ML17292B1321997-12-0404 December 1997 Proposed Tech Specs Modifying Min Critical Power Ratio Safety Limits ML17292B0281997-08-14014 August 1997 Proposed Tech Specs Revising TS 5.5.6 by Adding Note That Would Extend Interval Requirement to Perform Full Stroke Exercise Testing of TIP-V-6 Until 1998 Refueling Outage ML17292A9691997-08-12012 August 1997 Proposed Tech Specs Supporting Request for Enforcement Discretion for Period of 45 Days from TS Action 3.6.1.3.A Required Actions to Isolate Purge Line & Verify Penetration Flow Path Isolated Every 31 Days ML17292A9421997-07-16016 July 1997 Proposed Tech Specs Adding New Min Reactor Vessel Pressure Versus Reactor Vessel Metal Temp (P/T) Curves,Applicable Up to 12 EFPYs ML17292A8901997-06-0606 June 1997 Revised Tech Spec Page 2.0-1 Modified to Indicate That SLMCPR for ATRIUM-9X Fuel Applies Only to Cycle 13 & Corresponding Bases Pages ML17292A8651997-05-20020 May 1997 Proposed Tech Specs,Requesting Mod of Minimum Critical Power Ratio Safety Limits by 970615 ML17292A7631997-03-24024 March 1997 Rev 7 to Licensee Controlled Specs. ML17292A7621997-03-24024 March 1997 Rev 5 to TS Bases. ML17292A7581997-03-22022 March 1997 Proposed Tech Specs Modifying Response Time Testing SR for RPS Instrumentation,Primary Containment Isolation Actuation Instrumentation & ECCS Actuation Instrumentation ML17292A7531997-03-20020 March 1997 Proposed Tech Specs Re Response Time Testing Requirements ML17292A6591997-01-14014 January 1997 Proposed Tech Specs Reflecting Compilation of TS Change Requests Submitted to NRC in Ltrs Dtd 951208,960709 & 1212 ML17292A6341996-12-12012 December 1996 Proposed Tech Specs Requesting Conversion Based Upon NUREG-1434,rev 1 ML17292A5511996-10-15015 October 1996 Proposed Tech Specs Re Secondary Containment & SGTS to Reflect Revised Secondary Containment Drawdown & post- Accident Analyses Results ML17292A5411996-10-10010 October 1996 Proposed Tech Specs Requesting Addition of Section 2B(6) Re Storage of Byproduct,Source & Special Nuclear Materials ML17292A4501996-09-0606 September 1996 Proposed Tech Specs,Containing Corrections to Factual Statements & Proposed Info to Clarify Evaluations ML17292A4111996-08-0909 August 1996 Proposed Tech Specs,Revising TS Section 6.3 Re Unit Staff Qualifications,By Changing Operations Manager Qualification Requirements Associated W/Operations Knowledge from Meeting Ansi/Ans N18.1-1971 ML17292A3561996-07-0909 July 1996 Proposed Tech Specs,Revising Rev a, Including Changes in Vol 7.Proposed Rev Does Not Change Conclusion of NSHC or Environ Assessment Provided Rev a ML20107M3391996-04-24024 April 1996 Proposed Tech Specs,Modifying TS to Support Cycle 12, Scheduled to Begin Subsequent to Spring 1996 Outage ML17292A1511996-04-22022 April 1996 Proposed Tech Specs,Supplementing TS That Describes Administrative & Editorial Changes to Section 6.0, Administrative Controls. ML17291B2801996-03-19019 March 1996 Proposed Tech Specs Re Containment Leakage Testing ML17291B2491996-02-26026 February 1996 Proposed Tech Specs,Submitting Revised Copy of TS Bases Which Include Minor Changes & Clarifications Made Per Requirements of 10CFR50.59 ML17333A0201996-01-19019 January 1996 Proposed Tech Specs Re Primary Containment Leakage Testing ML17291B0941995-10-26026 October 1995 Proposed Tech Specs,Replacing Existing Reactor Recirculation Flow Control Sys W/Adjustable Speed Drive Sys ML17291A9911995-08-16016 August 1995 Proposed Tech Specs Page 3/4 4-4,incorporating Surveillance Notes in Front of Surveillances 4.4.1.2.1 & 4.4.1.2.2 for Jet Pump Operability to Clarify That Notes Apply to Each Surveillance ML17291A8441995-06-0606 June 1995 Proposed Tech Specs Section 6.0, Administrative Controls. ML17291A8401995-06-0606 June 1995 Proposed Tech Specs Index,Deleting Ref to Bases Pages ML17291A8371995-06-0606 June 1995 Proposed Tech Specs Section 6.9.3.2,adding Ref to Three TRs Describing Analytical Methods That May Be Used in Determining Reactor Core Operating Limits for Reload Licensing Applications ML17291A7561995-04-25025 April 1995 Proposed Tech Specs,Adding RWCU Sys High Blowdown Containment Isolation Trip Function & Associated LCO & SRs to Tables 3.3.2-1,3.3.2-2 & 4.3.2.1-1 ML17291A6541995-02-10010 February 1995 Proposed Tech Specs,Modifying Surveillance Acceptance Criteria from 10% to 20% for Individual Jet Pump diffuser- to-lower Plenum Differential Pressure Variations of Individual Jet Pump from Established Patterns ML17291A4811994-10-31031 October 1994 Proposed Tech Spec Relocating Safety/Relief Valve Position Indication Instrumentation Requirements ML17291A4781994-10-31031 October 1994 Proposed Tech Spec 3/4.1.3.1, Reactivity Control Sys. ML17291A4451994-10-12012 October 1994 Corrected Proposed TS Bases 3/4.2.6, Power/Flow Instability. ML17291A4221994-09-26026 September 1994 Proposed Tech Specs,Reflecting Use of Siemens Power Corp Staif Code for Stability Analysis,Per Ieb 88-007,Suppl 1 ML17291A3981994-09-18018 September 1994 Proposed TS Table 3.6.3-1 Re Primary Containment Isolation Valve Requirements ML17291A3191994-08-0808 August 1994 Proposed Tech Specs 4.0.5 Re Guideliness for Inservice Insp & Testing Program ML17291A2171994-07-12012 July 1994 Proposed Tech Specs for Relocation of TS Tables for Instrument Response Time Limits ML17291A2221994-07-0808 July 1994 Proposed TS W/Regard to Control Rod Scram Insertion Testing Under Emergency Circumstances ML17291A1561994-06-23023 June 1994 Proposed Tech Specs Re Supporting Hydrostatic Testing 1999-07-29
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML17284A9121999-10-13013 October 1999 Proposed Tech Specs 3.3.6.1,Table 3.3. 6.1-1, Primary Containment Isolation Instrumentation. ML17284A8761999-08-27027 August 1999 Replacement Page 9 of 9 to Attachment 4 of Procedure 13.10.6 ML17284A8491999-07-29029 July 1999 Proposed Tech Specs,Revising SR 3.5.2.2 Re Condensate Storage Tank Water Level ML17284A8421999-07-29029 July 1999 Proposed Tech Specs Revising SR of TS 3.8.4, DC Sources - Operating & SR 3.8.5.1 of TS 3.8.5, DC Sources - Shutdown. ML17284A8461999-07-29029 July 1999 Proposed Tech Specs,Revising Table 3.3.5.1-1, ECCS Instrumentation Items 1.a,2.a,4.a & E.A. ML17284A8521999-07-29029 July 1999 Proposed Tech Specs 3.4.9, RHR Shutdown Cooling Sys - Hot Shutdown. ML17333A0021999-04-20020 April 1999 Proposed Tech Specs Section 3.4.11,replacing Existing Reactor Pressure Temp Limit Curves by 000630 ML17292B6341999-04-0707 April 1999 Proposed Tech Specs Modifying MCPR Safety Limits to Allow Continued Power Operation at Plant Following Restart from R-14 RFO ML17292B5731999-03-0101 March 1999 ODCM for WNP-2 ML17292B4881998-12-17017 December 1998 Proposed Tech Specs SR 3.8.1.8,allowing Capability to Manually Transfer Between Preferred & Alternate Offsite Power Sources During Modes 1 & 2 by 990125 ML20198A7051998-11-30030 November 1998 Revs 8 Through 13 to TS Bases & Revs 12 Through 15 of Licensee Controlled Specs ML17284A7181998-08-0505 August 1998 Proposed Tech Specs SR 3.8.4.7,allowing Performance Discharge Test to Be Performed in Lieu of Battery Svc Test for Div 1 & 3,125 Vdc Batteries E-B1-1 & HPCS-B1-DG3 & Div 1,250 Vdc Battery E-B2-1 ML17284A7071998-07-17017 July 1998 Proposed Tech Specs Modifying SR 3.8.4.7 to Allow Performance Discharge Test to Be Performed in Lieu of Battery Svc Test for Div 2,125 Vdc,Battery E-B1-2 ML17284A6431998-05-29029 May 1998 Revised Plant Procedure Sys for Site Wide Procedures, Replacing Pages Located in Manual W/Pages in Package ML17292B2831998-03-0909 March 1998 Proposed Tech Specs Establishing Interim SLMCPR for Siemens Power Corp ATRIUM-9X Fuel Applicable to Cycle 14 Only ML17292B2591998-01-31031 January 1998 Offsite Dose Calculation Manual. ML17292B1321997-12-0404 December 1997 Proposed Tech Specs Modifying Min Critical Power Ratio Safety Limits ML17292B0281997-08-14014 August 1997 Proposed Tech Specs Revising TS 5.5.6 by Adding Note That Would Extend Interval Requirement to Perform Full Stroke Exercise Testing of TIP-V-6 Until 1998 Refueling Outage ML17292A9691997-08-12012 August 1997 Proposed Tech Specs Supporting Request for Enforcement Discretion for Period of 45 Days from TS Action 3.6.1.3.A Required Actions to Isolate Purge Line & Verify Penetration Flow Path Isolated Every 31 Days ML17292A9421997-07-16016 July 1997 Proposed Tech Specs Adding New Min Reactor Vessel Pressure Versus Reactor Vessel Metal Temp (P/T) Curves,Applicable Up to 12 EFPYs ML17292A8901997-06-0606 June 1997 Revised Tech Spec Page 2.0-1 Modified to Indicate That SLMCPR for ATRIUM-9X Fuel Applies Only to Cycle 13 & Corresponding Bases Pages ML17292A8651997-05-20020 May 1997 Proposed Tech Specs,Requesting Mod of Minimum Critical Power Ratio Safety Limits by 970615 ML17292A8301997-03-31031 March 1997 Wppss WNP-2 RPV Surveillance Matls Testing & Analysis. ML17292A7621997-03-24024 March 1997 Rev 5 to TS Bases. ML17292A7631997-03-24024 March 1997 Rev 7 to Licensee Controlled Specs. ML17292A7581997-03-22022 March 1997 Proposed Tech Specs Modifying Response Time Testing SR for RPS Instrumentation,Primary Containment Isolation Actuation Instrumentation & ECCS Actuation Instrumentation ML17292A7531997-03-20020 March 1997 Proposed Tech Specs Re Response Time Testing Requirements ML17292A6591997-01-14014 January 1997 Proposed Tech Specs Reflecting Compilation of TS Change Requests Submitted to NRC in Ltrs Dtd 951208,960709 & 1212 ML17292A6341996-12-12012 December 1996 Proposed Tech Specs Requesting Conversion Based Upon NUREG-1434,rev 1 ML17292A6161996-11-19019 November 1996 Rev 1 to WNP-2 IST Program Plan (Pumps & Valves) 2nd Interval (941213-041212). ML17292A5511996-10-15015 October 1996 Proposed Tech Specs Re Secondary Containment & SGTS to Reflect Revised Secondary Containment Drawdown & post- Accident Analyses Results ML17292A5411996-10-10010 October 1996 Proposed Tech Specs Requesting Addition of Section 2B(6) Re Storage of Byproduct,Source & Special Nuclear Materials ML17292A4501996-09-0606 September 1996 Proposed Tech Specs,Containing Corrections to Factual Statements & Proposed Info to Clarify Evaluations ML17292A4111996-08-0909 August 1996 Proposed Tech Specs,Revising TS Section 6.3 Re Unit Staff Qualifications,By Changing Operations Manager Qualification Requirements Associated W/Operations Knowledge from Meeting Ansi/Ans N18.1-1971 ML17292A3561996-07-0909 July 1996 Proposed Tech Specs,Revising Rev a, Including Changes in Vol 7.Proposed Rev Does Not Change Conclusion of NSHC or Environ Assessment Provided Rev a ML17292A7241996-05-31031 May 1996 Offsite Dose Calculation Manual. ML17292A2741996-04-25025 April 1996 Rev 0 to UT-WNP2-208V0, Exam Summary Sheet. ML20107M3391996-04-24024 April 1996 Proposed Tech Specs,Modifying TS to Support Cycle 12, Scheduled to Begin Subsequent to Spring 1996 Outage ML17292A1511996-04-22022 April 1996 Proposed Tech Specs,Supplementing TS That Describes Administrative & Editorial Changes to Section 6.0, Administrative Controls. ML17291B2801996-03-19019 March 1996 Proposed Tech Specs Re Containment Leakage Testing ML17291B2491996-02-26026 February 1996 Proposed Tech Specs,Submitting Revised Copy of TS Bases Which Include Minor Changes & Clarifications Made Per Requirements of 10CFR50.59 ML17333A0201996-01-19019 January 1996 Proposed Tech Specs Re Primary Containment Leakage Testing ML17291B1751995-12-31031 December 1995 Reactor Power Uprate Startup Test Rept, for WNP-2. W/951215 Ltr ML17291B0941995-10-26026 October 1995 Proposed Tech Specs,Replacing Existing Reactor Recirculation Flow Control Sys W/Adjustable Speed Drive Sys ML17291A9911995-08-16016 August 1995 Proposed Tech Specs Page 3/4 4-4,incorporating Surveillance Notes in Front of Surveillances 4.4.1.2.1 & 4.4.1.2.2 for Jet Pump Operability to Clarify That Notes Apply to Each Surveillance ML17291A9591995-07-28028 July 1995 Operations Instructions OI-23,Rev a to, Human Performance Improvement Program. ML20087E2831995-07-26026 July 1995 Performance Enhancement Strategy 1995 ML17291A8401995-06-0606 June 1995 Proposed Tech Specs Index,Deleting Ref to Bases Pages ML17291A8371995-06-0606 June 1995 Proposed Tech Specs Section 6.9.3.2,adding Ref to Three TRs Describing Analytical Methods That May Be Used in Determining Reactor Core Operating Limits for Reload Licensing Applications ML17291A8441995-06-0606 June 1995 Proposed Tech Specs Section 6.0, Administrative Controls. 1999-08-27
[Table view] |
Text
TABLE 3.6. 3-1 Continued PRIMARY CONTAINMENT ISOLATION VALVES CI MAXIMUM ISOLATION TIME VALVE FUNCTION AND NUMBER Seconds n
I Pl Automatic Isolation Valves (Continued)
Bl I Equipment Drain-(Radioactive)
EOR-V-19 hh
~ EDR-V-20 Floor Drain (Radioactive) r4 15 FOR-V-3 FDR-V-4 D
Fuel Pool Cooling/Suppression Pool Cleanup 35 Q0<
,,a>
xa FPC-V-149 FPC-V" 153(f)
- pm
, FPC-V-154(f)
H~ FPC"V-156 Reactor Recirculation ttydraulic Control(e) (g) 15 l
- 'o ON Ul>
ttY-V-liA,B so~ tlY-V-lBA,B llY-V-19A,B IlY-V-20A,B tlY"V"33A,B llY-V-34A,B ttY-V-35A,B tlY-V-36A,B O
Traversing Incore Probe TIP-V-1,2;3,4,5
PRIMARY CONTAINMENT ISOLATION VALVFS HAXIHUH ISOLATION TINE VALVE FUNCTION AND NUMBER VALV GROUP a Seconds
- d. Other Containment Isolation Valves (Continued)
Radiation Honitoring N.A.
P I-V-X72 fI I PI-V-X73e/I Transversing Incore Probe System N.A.
TIP-V-6 TIP-V-7,8,9,10,11(e)
TABLE NOTATIONS
- But greater than 3 seconds.
OProvisions of Technical Specification 3.0.4 are not applicable.
(a) See Technical Specification 3.3.2 for the isolation signal(s) which operate each group.
(b) Valve leakage not included in sum of Type B and C tests.
(c) Hay be opened on an intermittent basis under administrative cont,rol.
(d) Not closed by SLC actuation signal.
(e) Not subject to Type C Leak Rate Test.
(f) Hydraulic leak test at 38.2 psig.
(9) Not subject to Type C test. Test per Technical Specification 4.4.3.2.2 (h) Tested as part of Type A test.
(i) Hay be tested as part of Type A test. If so tested, Type C test results pay be excluded from sum of other Type B and C tests.
(j) Reflects closure times for containment isolation only.
(k) Ouring operational conditions 1, 2 8 3 the requirement for automatic isolation does not apply to RHR-V-8.
Except that RHR-V-8 may be opened in operational conditions 2 8 3 provided control is returned to the control room, with the interlocks reestablished, and reactor pressure is less than 135 psig.
The isolation logic associated with the reactor recirculation hydraulic control containment isolation valves need not meet single failure 1995.
criteria for OPERABILITY for a period ending no later than Hay 15,
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nee 12AX 12AY nee 13AY HYX1-A 12AX N-V-1T,18 19.28 Cir cuit normally ener gized 12AY Shown in De-Ener gized Condition HY V-3330 35@
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Attachment SAFETY ASSESSMENT OF HYDRAULICLINE FAILURE VS SAFETY ASSESSMENT OF POTENTIAL CORRECTIVE ACTIONS This analysis compares the increased risk from continued operation with the current containment isolation logic design (potential containment bypass path) with the increase in risk associated with possible corrective actions. The containment bypass scenario is presented in terms of containment'ailure probability coincident with core damage from a large LOCA while the corrective actions are presented in terms of increased core damage frequency. The bypass scenario requires a LOCA and a sequence of equipment failures, the corrective actions involve somewhat more probable events such as loss of feedwater or turbine trip.
n inment 8 ass cenario Lack of two completely independent, automatically actuated isolation valves on the Recirculation Flow Control Valve (FCV) hydraulic lines implies that during a demand for containment isolation these hydraulic lines could present a direct path from containment to the environment.
Situation: The recirculation FCV hydraulic lines are designed for an internal pressure of 2200 psig and the system is designed and constructed to operate as a closed loop:
Each of the two FCV Actuator Hydraulic Power Units, HY-HP-3A and -B is installed outside containment Supply to the actuators, installed inside containment, is maintained at 1800 psig so the "pre-accident" integrity of the piping is confirmed Isolation of the hydraulic supply lines is achieved with two series isolation valves isolation of each of the four sets of two series valves in line "A" is dependent upon Div "1" actuation isolation of each of the four sets of two series valves in line "B" is dependent upon Div "2" actuation valves are functionally tested each refueling Calculation of Conditional Containment Failure Probability:
o Initiating event: Large LOCA (3E-4/yr) o Probability that Large LOCA originates with Recirculation pump discharge piping = 0.1 (based on estimate that recirculation pump discharge piping represents 10% of large in-containment piping)
Attachment SAFETY ASSESSMENT OF HYDRAULICLINE FAILURE VS SAFETY ASSESSMENT OF POTENTIAL CORRECTIVE ACTIONS o Conditional failure probability of hydraulic piping inside containment given failure of recirculation piping = 1.0 Note: Failure is assumed to result from movement of the pump discharge piping initiating a hydraulic line failure at the actuator - no credit is given for the ameliorating effects of the flexible hose connections between the valve actuators and the hydraulic lines inside containment. There are actually no postulated breaks that would cause failure of the hydraulic lines.
Probability that the hydraulic lines will fail outside containment is assumed to be 1.0 (a factor of 1E-2 could be justified for equivalent instrument line break accidents). There are actually no postulated breaks that would cause failure of the hydraulic lines.
Probability that one of the two actuation systems fails on demand = 2 ":
8.3E-3 Note: calculated as follows: 1.27E-6/hr' 13,140 hrs between tests (18 mo test interval) 1/2 (finds average probability over the interval)
'elay failure rate taken from NPRDS, NUREG/CR-2815 gives 1E-6 per hour.
o Conditional probability of core damage given a Large LOCA = 1E-4 The annual calculated frequency for this containment bypass scenario coincident with core damage from a large LOCA is 5E-11 per year (less than the Individual Plant Evaluation (IPE) truncation value of 1E-9/yr).
Other LOCA core melt scenarios may be more frequent, but in other scenarios the conditional probability of hydraulic line failure will be lower because there will be less dependency between the initiating event and induced failure of the piping. Based on the missile hazards analysis performed for piping inside containment, the hydraulic lines are not a target of any LOCA originated missiles or jets and so will not fail inside containment as a result of the large LOCA.
Additionally, the lines are small (( 3/4"), releases through these lines will be much smaller than any considered previously for equipment qualification purposes. Therefore, a sequence of events involving core damage, rupture of the lines and failure of ECCS equipment caused by steam and radiation releases through the open hydraulic lines is not considered to be credible.
0 A hmn SAFETY ASSESSMENT OF HYDRAULICLINE FAILURE VS SAFETY ASSESSMENT OF POTENTIAL CORRECTIVE ACTIONS Safe Assessment of orrective Actions Manual Shu d wn The WNP-2 IPE assumes 0.5 manual shutdowns per year. Based on this initiator frequency the total core damage frequency due to manual shutdown events is calculated to be 4.5E-8 per year.
Based on a sensitivity study using the WNP-2 IPE model, an increase in manual shutdown frequency to 1.5 per year increases its total contribution to core damage frequency to 15.8E-8 per year.
era ion with H draulic I la ion V lve I Isolation of the hydraulic lines (closure of the isolation valves) for the rest of this cycle can be used as a compensatory measure, however, this would result in loss of all recirculation flow control. As a result, the plant would be unable to respond to a relatively minor transient in the feedwater system. This means that a feedwater transient would initiate a plant SCRAM and has the potential for increasing risk.
Feedwater transients are typically encountered about 3 times per year, and as a result, plant trips could be expected to increase from a current level of 4 per year to 7 per year. This assumes that the plant would be unable to respond to even a minor transient. Based on a sensitivity study using the WNP-2 IPE model, even a single event increase resulting in a plant SCRAM will increase core damage frequency by approximately 3E-7 per year.
onclu ion Manual shutdown and/or isolation of the hydraulic lines is not recommended. The increase in risk from the potential corrective actions is greater than the risk of allowing plant operation in the current configuration.
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