ML042390204
| ML042390204 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 08/20/2004 |
| From: | Lucas J Progress Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RNP-RA/04-0067 | |
| Download: ML042390204 (23) | |
Text
10 CFR 50.90 Progress Energy Serial: RNP-RA/04-0067 AUG 2 0 2004 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGES REGARDING REACTOR PROTECTON SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TABLES Ladies and Gentlemen:
In accordance with the provisions of 10 CFR 50.90, Progress Energy Carolinas, Inc. (PEC), also known as Carolina Power and Light Company, is submitting a request for an amendment to the Technical Specifications (TS) contained in Appendix A of the Operating License for H. B.
Robinson Steam Electric Plant (HBRSEP), Unit No. 2.
The proposed amendment revises the Allowable Values for the following Reactor Protection System (RPS) instrumentation functions: Intermediate Range Neutron Flux, Reactor Coolant Flow - Low, Steam Generator (SG) Water Level - Low Coincident with Steam Flow/Feedwater Flow Mismatch, and Intermediate Range Neutron Flux (P-6) Interlock. Additionally, these changes revise the Allowable Value for the Engineered Safety Feature Actuation System Instrumentation function for High Steam Flow in Two Steam Lines Coincident with Steam Line Pressure - Low. Also included is the proposed deletion of an unnecessary footnote associated with the applicability for the Automatic Trip Logic RPS instrumentation function.
Attachment I provides an Affirmation as required by 10 CFR 50.30(b).
Attachment II provides a description of the current condition, a description and justification of the proposed change, a No Significant Hazards Consideration Determination, and an Environmental Impact Consideration.
Attachment III provides a markup of the affected TS pages.
Attachment IV provides a retyped version of the affected TS pages.
Attachment V provides calculations in support of the proposed Allowable Value changes.
Progress Energy Carolinas, Inc.
Robinson Nuclear Plant 3581 West Entrance Road Hartsville, SC 29550 A
United States Nuclear Regulatory Commission Serial: RNP-RA/04-0067 Page 2 of 2 Attachment VI provides the setpoint methodology procedure used in the determination of the proposed Allowable Value changes.
In accordance with 10 CFR 50.91(b), PEC is providing the State of South Carolina with a copy of this license amendment request.
PEC requests approval of this license amendment request by May 15, 2005.
If you have any questions concerning this matter, please contact Mr. C. T. Baucom at (843) 857-1253.
Sincerely, Manager - Support Services - Nuclear Attachments:
I.
Affirmation II.
Request for Technical Specifications Changes Regarding Reactor Protection System and Engineered Safety Feature Actuation System Instrumentation Tables III.
Markup of Technical Specifications Pages IV.
Retyped Technical Specifications Pages V.
Supporting Calculations VI.
Setpoint Methodology Procedure CTB/cac c:
Mr. T. P. O'Kelley, Director, Bureau of Radiological Health (SC)
Mr. H. J. Porter, Director, Division of Radioactive Waste Management (SC)
Dr. W. D. Travers, NRC, Region II Mr. C. P. Patel, NRC, NRR NRC Resident Inspector, HBRSEP Attorney General (SC)
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/04-0067 Page 1 of 1 AFFIRMATION The information contained in letter RNP-RA/04-0067 is true and correct to the best of my information, knowledge, and belief; and the sources of my information are officers, employees, contractors, and agents of Progress Energy Carolinas, Inc., also known as Carolina Power and Light Company. I declare under penalty of perjury that the foregoing is true and correct.
Executed On:,O
. Adz l
l VSJ.
W.M t
/er VW resident, H
~SEP. Unit No. 2
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0067 Page 1 of 7 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGES REGARDING REACTOR PROTECTION SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENATION TABLES Description of Current Condition Appendix A, Technical Specifications (TS), to Operating License (OL) No. DPR-23, for H. B.
Robinson Steam Electric Plant (HBRSEP), Unit No. 2, establishes the Limiting Condition for Operation (LCO) requirements for the following Protection Functions in Section 3.3:
- TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 3, Intermediate Range Neutron Flux.
- TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 9, Reactor Coolant Flow - Low.
- TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 14, Steam Generator (SG) Water Level - Low Coincident with Steam Flow - Feedwater Flow Mismatch.
- TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 17.a, Intermediate Range Neutron Flux (P-6) Interlock.
- TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 20, Automatic Trip Logic.
- TS Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, Function 1.g, High Steam Flow in Two Steam Lines Coincident with Steam Line Pressure - Low.
Description and Justification of the Proposed Changes The proposed changes associated with the Allowable Values for the functions listed in TS Section 3.3.1 are consistent with the TS Bases, which state:
"Setpoints in accordance with the Allowable Value ensure that [safety limits] SLs are not violated during [anticipated operational occurrences] AOOs (and that the consequences of
[design basis accidents] DBAs will be acceptable, providing the unit is operated from within the LCOs at the onset of the AOO or DBA and the equipment functions as designed). Note that in the accompanying LCO 3.3.1, the Allowable Values are the [limiting safety system settings] LSSS."
The Bases for TS Section 3.3.1 further state:
"The Nominal Trip Setpoints and Allowable Values listed in Table 3.3.1-1 are based on the methodology described in the company setpoint methodology procedure (Ref. 8), which incorporates all of the applicable uncertainties for each channel. The magnitudes of these uncertainties are factored into the determination of each Nominal Trip Setpoint. All field sensors and signal processing equipment for these channels are assumed to operate within the allowances of these uncertainty magnitudes."
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0067 Page 2 of 7 The proposed change associated with the Allowable Value for the function listed in TS Section 3.3.2 is consistent with the TS Bases, which state:
"Setpoints in accordance with the Allowable Value ensure that the consequences of Design Basis Accidents (DBAs) will be acceptable, providing the unit is operated from within the LCOs at the onset of the DBA and the equipment functions as designed."
The Bases for TS Section 3.3.2 further state:
"The Nominal Trip setpoints and Allowable Values listed in Table 3.3.2-1, are based on the methodology described in the company setpoint methodology procedure (Ref. 9), which incorporates all of the applicable uncertainties for each channel. The magnitudes of these uncertainties are factored into the determination of each Nominal Trip Setpoint. All field sensors and signal processing equipment for these channels are assumed to operate within the allowances of these uncertainty magnitudes."
The company setpoint methodology procedure (listed as Ref. 8 and 9 in the TS Bases excerpt paragraphs) is listed in the HBRSEP, Unit No. 2, TS as "Attachment VII to CP&L's letter to NRC dated May 30, 1997, 'H. B. Robinson Steam Electric Plant, Unit No. 2, Response to Request for Additional Information Regarding the Technical Specifications Change Request to Convert to the Improved Standard Technical Specifications."' The current methods being used for HBRSEP, Unit No. 2, setpoint calculations remain consistent with this reference and are contained in Progress Energy Nuclear Generation Group Procedure, EGR-NGGC-0153, "Engineering Instrument Setpoints." A copy of this procedure is provided in Attachment VI to this letter.
Section 9.1.1 of procedure EGR-NGGC-0153 states the following:
"Application of the methodology described in this procedure is appropriate for Limiting Safety System Settings as defined in 10 CFR 50.36, and for operator indications when required by the emergency response guidelines. Where Limiting Safety System Settings have been established for nuclear plant instruments by the plant Technical Specifications, the settings are to be chosen so that automatic protective action will occur to protect against the most severe abnormal situation without exceeding analytical safety limits. Instruments that are utilized to ensure that these safety limits are not exceeded will provide adequate margins to safety which are to be documented through the use of instrument uncertainty and scaling calculations."
The methodology used in the determination of the proposed Allowable Value changes is described in EGR-NGGC-0153, "Engineering Instrument Setpoints." A specific description of and justification for each of these proposed changes is provided as follows:
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0067 Page 3 of 7 Intermediate Range Neutron Flux TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 3, Intermediate Range Neutron Flux, is being revised to change the Allowable Value from < 37.02% to < 36.40%, and TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 17.a, Intermediate Range Neutron Flux (P-6) Interlock, is being revised to change the Allowable Value from > 7.29 E-1 1 amps to > 9.34 E-1 1 amps. These changes to the Allowable Values for these functions are based on the calculation RNP-I/INST-1 135, "Nuclear Instrumentation Intermediate Range Error Analysis."
A copy of that calculation is provided in Attachment V to this letter.
The proposed changes to these Allowable Values provide more restrictive limits for these parameters. Specifically, the proposed revision to the Allowable Value for the Intermediate Range Neutron Flux trip from < 37.02% to < 36.40% reduces the operability limit range for this trip function. Similarly, the increase in the Intermediate Range Neutron Flux (P-6) Interlock Allowable Value from > 7.29 E-1 1 amps to > 9.34 E-1 1 amps reduces the operability limit range for this interlock.
The proposed change to the Allowable Value for the Intermediate Range Neutron Flux trip reactor protection function will ensure that the operability limit for this function is properly established.
Therefore, the consequences of Design Basis Accidents (DBAs) will be acceptable, providing the unit is operated from within the LCOs at the onset of the DBA, and the equipment functions as designed.
Reactor Coolant Flow - Low TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 9, Reactor Coolant Flow -
Low, is being revised to change the Allowable Value for both the "Single Loop" and "Two Loops" sub-items from > 93.47% to > 93.45%. These changes to the Allowable Values for these functions are based on the calculation RNP-I/INST-1 128, "RCS Flow Instrument Uncertainty and Scaling Calculation." A copy of that calculation is provided in Attachment V to this letter.
The proposed changes to these Allowable Values provide slightly less restrictive limits for these parameters. Specifically, the proposed revisions to the Allowable Values for the Reactor Coolant Flow - Low RPS instrumentation functions slightly increase the operability limit range for these trip functions.
The proposed changes to the Allowable Values for the Reactor Coolant Flow - Low reactor protection functions will ensure that the operability limits for these functions are properly established. Therefore, the consequences of DBAs will be acceptable, providing the unit is operated from within the LCOs at the onset of the DBA, and the equipment functions as designed.
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0067 Page 4 of 7 Steam Generator Water Level - Low TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 14, SG Water Level - Low Coincident with Steam Flow - Feedwater Flow Mismatch, is being revised to change the Allowable Value from < 7.06 E5 Ibm/hr to < 7.01 E5 Ibm/hr. This change to the Allowable Value is based on the calculation RNP-I/INST-1041, "Feedwater Flow Loop Uncertainty and Scaling Calculation." A copy of that calculation is provided in Attachment V to this letter.
The proposed change to this Allowable Value provides a slightly more restrictive limit for this parameter. Specifically, the proposed revision to the Allowable Value for the SG Water Level -
Low Coincident with Steam Flow - Feedwater Flow Mismatch reactor protection function slightly decreases the operability limit range for this trip function The proposed change to the Allowable Value for the SG Water Level - Low Coincident with Steam Flow - Feedwater Flow Mismatch reactor protection function will ensure that the operability limit for this function is properly established. Therefore, the consequences of DBAs will be acceptable, providing the unit is operated from within the LCOs at the onset of the DBA, and the equipment functions as designed.
Automatic Trip Logic TS Table 3.3.1-1, Reactor Protection System Instrumentation, Function 20, Automatic Trip Logic, is being revised to remove an unnecessary note associated with the Applicable Modes for this function. Note "j" is associated with MODE 1 for this function. Note "j" states, "Below the P-6 (Intermediate Range Neutron Flux) interlock for the logic inputs from Source Range Neutron Flux detector channels." This note is not needed, because the Automatic Trip Logic function is only required to be operable when the associated reactor protection functions are required to be operable.
This change also revises the applicability for this LCO to be consistent with the current version of NUREG-1431, "Improved Standard Technical Specifications for Westinghouse Plants."
High Steam Line Flow in Two Steam Lines TS Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, Function 1.g, High Steam Flow in Two Steam Lines Coincident with Steam Line Pressure - Low, is being revised to change the Allowable Value from > 605.05 psig to > 597.76 psig. This change to the Allowable Value is based on the calculation RNP-I/INST-1043, "Main Steam Pressure Uncertainty and Scaling Calculation." A copy of that calculation is provided in Attachment V to this letter.
The proposed change to the Allowable Value for the High Steam Flow in Two Steam Lines Coincident with Steam Line Pressure - Low engineered safety feature safety injection (SI) actuation function is a less restrictive change based on the results of the associated setpoint calculation. The proposed Allowable Value will continue to provide an appropriate operability limit for this SI actuation function.
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0067 Page 5 of 7 No Significant Hazards Consideration Determination Progress Energy Carolinas, Inc. (PEC), also known as Carolina Power and Light Company, is proposing changes to Appendix A, Technical Specifications, of Facility Operating License No.
DPR-23, for the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. The proposed changes revise the Allowable Values for the following Reactor Protection System (RPS) instrumentation functions: Intermediate Range Neutron Flux, Reactor Coolant Flow - Low, Steam Generator (SG) Water Level - Low Coincident with Steam Flow/Feedwater Flow Mismatch, and Intermediate Range Neutron Flux (P-6) Interlock. Additionally, these changes revise the Allowable Value for the Engineered Safety Feature Actuation System Instrumentation function for High Steam Flow in Two Steam Lines Coincident with Steam Line Pressure - Low. Also included is the proposed deletion of an unnecessary footnote associated with the applicability for the Automatic Trip Logic RPS instrumentation function.
An evaluation of the proposed changes has been performed in accordance with 10 CFR 50.91(a)(1) regarding no significant hazards considerations using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this amendment request follows:
- 1.
Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposal to revise the Allowable Values for the affected reactor protection and engineered safety feature actuation functions was developed in accordance with the current setpoint methodology for HBRSEP, Unit No. 2, thus ensuring that the probability and consequences of previously evaluated accidents are not significantly increased. The proposed deletion of the unnecessary footnote associated with the Automatic Trip Logic reactor protection instrumentation function does not change the requirements for operability of this function. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated, because the factors that are used to determine the probability and consequences of accidents are not being affected.
- 2.
Do the proposed changes create the possibility of a new or different kind of accident from any previously evaluated?
The proposed changes will continue to ensure that the operability of the previously described functions will be appropriately maintained. No physical changes to the HBRSEP, Unit No. 2, systems, structures, or components are being implemented. There are no new or different accident initiators or sequences being created by the proposed Technical Specifications changes. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/04-0067 Page 6 of 7
- 3.
Do the proposed changes involve a significant reduction in the margin of safety?
The proposed changes, as previously described, ensure that the margin of safety for the applicable fission product barriers that are protected by these functions will continue to be maintained. This conclusion is based on the use of a valid setpoint methodology for determining the Allowable Values for the reactor protection and engineered safety feature actuation functions. Therefore, these changes do not involve a significant reduction in the margin of safety.
Based on the preceding discussion, the requested changes do not involve a significant hazards consideration.
Environmental Impact Consideration 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions for categorical exclusion for performing an environmental assessment. A proposed change for an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed change would not (i) involve a significant hazards consideration; (ii) result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite; (iii) result in a significant increase in individual or cumulative occupational radiation exposure. PEC has reviewed this request and determined that the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the proposed amendment. The basis for this determination follows.
Proposed Change Progress Energy Carolinas, Inc. (PEC), also known as Carolina Power and Light Company, is proposing changes to Appendix A, Technical Specifications, of Facility Operating License No. DPR-23, for the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. The proposed changes revise the Allowable Values for the following Reactor Protection System (RPS) instrumentation functions: Intermediate Range Neutron Flux, Reactor Coolant Flow - Low, Steam Generator (SG) Water Level - Low Coincident with Steam Flow/Feedwater Flow Mismatch, and Intermediate Range Neutron Flux (P-6) Interlock. Additionally, these changes revise the Allowable Value for the Engineered Safety Feature Actuation System Instrumentation function for High Steam Flow in Two Steam Lines Coincident with Steam Line Pressure - Low. Also included is the proposed deletion of an unnecessary footnote associated with the applicability for the Automatic Trip Logic RPS instrumentation function.
United States Nuclear Regulatory Commission 1 to Serial: RNP-RA/04-0067 Page 7 of 7 Basis The proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:
- 1.
As demonstrated in the No Significant Hazards Consideration Determination, the proposed changes do not involve a significant hazards consideration.
- 2.
The proposed changes revise the Allowable Values for the following RPS instrumentation functions: Intermediate Range Neutron Flux, Reactor Coolant Flow - Low, Steam Generator (SG) Water Level - Low Coincident with Steam Flow - Feedwater Flow Mismatch, and Intermediate Range Neutron Flux (P-6) Interlock. Additionally, these changes revise the Allowable Value for the Engineered Safety Feature Actuation System instrumentation function for High Steam Flow in Two Steam Lines Coincident with Steam Line Pressure -
Low. Also included is the proposed deletion of an unnecessary footnote associated with the applicability for the Automatic Trip Logic RPS instrumentation function. These changes do not affect the generation or control of effluents. Therefore, the proposed changes will not result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite.
- 3.
The proposed changes, as previously described, do not affect any parameters that would cause an increase in occupational radiation exposure. There are no proposed physical changes to the facility or facility processes that would result in increased radiation exposure to plant personnel. Therefore, the proposed changes will not result in a significant increase in individual or cumulative occupational radiation exposure.
United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/04-0067 6 Pages (including cover page)
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGES REGARDING REACTOR PROTECTION SYSTEM AND ENGINEERED SAFETY lEATURE ACTUATION SYSTEM INSTRUMENTATION TABLES MARKUP OF TECHNICAL SPECIFICATIONS PAGES
RPS Instrumentation 3.3.1 Table 3.3.1*1 (page 1 of 7)
Reactor Protection System Instrumentation APPLICABLE MODES NOMINAL OR OTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE SETPOINT FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREIIENTS VALUE (1)
- 1. Manual Reactor Trip 1.2 3(a). 4 (a) 5(a) 2 2
B SR 3.3.1.14 C
SR 3.3.1.14 NA NA NA NA
- 2.
Power Range Neutron Flux
- a. High 1.2 4
0 SR 3.3.1.1 SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 4
E SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 z llO.93t lost RTP RTP (2)
- b.
Low
- 3.
Intermediate Range Neutron Flux (b). 2(0 2 (d)
- 4. Source Range Neutron Flux 2 (d) 2 F.G SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 2
H SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 2
IJ SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 2
L.K SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.11 1
L SR 3.3.1.1 SR 3.3.1.11 RTP RTP RTP a 1.28 ES 1.0 ES cps cps
(continued)
(1) A channel is OPERABLE with an actual Trip Setpoint value found outside Its calibration tolerance band provided the Trip Setpoint value Is conservative with respect to its associated Allowable Value and the channel Is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(2)
The Nominal Trip Setpoint Is as stated unless reduced as required by one or more of the following requirements:
LCO 3.2.1 Required Action A.2.2; LCO 3.2.2 Required Action A.1.2.2; or LCOD 3.7.1 Required Action 5.2.
(a)
With Pod Control System capable of rod withdrawal. or one or xore rods not fully inserted.
lb)
Below the P-10 (Power Range Neutron Flux) interlock.
tc)
Above the P-6 (Intermedidte Range Neutron flux) Interlock.
(d)
Below the P-6 (Intermediate Range Neutron Flux) Interlock.
Ce)
With the RTBs open.
In this condition. source range Function does not provide reactor trip but does provide indication and alarm.
HBRSEP Unit No. 2 3.3-13 Amendment No. 176
RPS Instrumentation 3.3.1 Table 3.3.1*1 (page 3 of 7)
Reactor Protection System Instrumentation APPLICABLE MODES NOMINAL OR OTHER lTIP SPECIFIED RECUIRED SURVEILLANCE ALLD10ZLE SETPOINT FUNCTIOH CONDITIONS C)WJNELS CONDITIONS RECUIREMENTS VALUE (1)
- 9.
Reactor Coolant Flow - Low
- a. Single Loop I(g) 3 per N
SR 3.3.1.1 94.26t loop SR 3.3.1.7 SR 3.3.1.10
- b.
Two Loops 1(h) 3 per H
94.261 loop SR 3.3.1.7 SR 3.3.1.10
- 10.
Reactor Coolant Pump (RCP) Breaker Position
- a. Single Loop I) per 0
- b.
Twn oopo (h)
Ipr H
- 11.
Undervoltage I(f)
I per H
SR 3.3.1.9 a 2959 V 3120 V RcPs bus SR 3.3.1.10
- 12. Undeefrequency 1({)
I per N
SR 3.3.1.10 a 57.84 58.2 Hz RCPs bus SR 3.3.1.14 Hz
- 13.
Steam 1.2 3 per SG E
SR 3.3.1.1 I 15.3Ct 16S Generator (SG)
SR 3.3.2.7 Water Level - Low SR 3.3.1.10 Low (continued)
(1) A channel is OrERABLE with on actual Trip Setpoint value found outside its callbratin tolerance bard provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Noinal Trip Setpoint.
(f)
Above the P-7 (Low Power Reactor Trips Block) Interlock.
Ig)
Above the P-8 (Power Range Neutron Flux) interlock.
(h)
Above the P-7 (Low Power Reactor Trips Block) Interlock and below the P-8 (Power Range Neutron Flux) Interlock.
HBRSEP Unit No. 2 3.3 15 Amendment No. 176
RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 4 of 7)
Reactor Protection System Instrumentation APPLICABLE MODES NCHIIIAL OR OTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLWABLE SETPOINT FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (1)
- 14.
SG Water Level - Low Coincident with Steam Fow/
Fooiator Flow His match
- 15.
- a. Low Auto Stop Oil Pressure
- b. Turbine Stop Valve Closure
- 16.
Safety Injection (SI)
Input from Engineered Safety Feature Actuatinn System (ESFAS) 1.2 2 per SG E
SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10
- 29.36t 30t 1.2 2 per SG E
SR 3.3.1 1 a
6.4 E3 SR 3.3.1.7 tbmhr ln/hr SR 3.3.1.10 3
P SR 3.3.1.10 a 40.87 SR 3.3.1.15 psig 2
P SR 3.3.1.15 NA 45 psig NA 1.2 2 trains Q
(1)
A channel is OPERABLE with an actual Trip Setpolnt value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to Its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(f)
Above the P-7 (Low Power Reactor Trips Block) interlock.
HBRSEP Unit No. 2 3.3 16 Amendment No. 176
RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 7)
Reactor Protect Iun System Instrumentation APPLICABLE MODES NOMINAL OR OTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOaABLE SETPOINT FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (1)
- 17.
Reactor Protection System Interlocks
- a. Intermediate 2(d) 2 S
SR 3.3.1.11 1 E-10 Range Neutron SR 3.3.1.13 E-11 amp amp Flux. P-6
- b.
Low Power I
Iper T
SR 3.3.2.13 NA NA Reactor Trips train SR 3.3.1.14 Block. P-7
- c.
Power Range 1
4 T
SR 3.3.1.11 s 42.94S 401 RTP Neutron Flux.
- d.
Power Range 1.2 4
S SR 3.3.1.11 a 7.06t I0t RTP Neutron Flux.
P-I0 12.941 RTP
- e. Turbine lp~ulse 1
2 T
- 10.711 101 Pressure.
.7 SR 3.3.1.10 turbine turbine Input SR 3.3.1.13 power power
- 18.
Reactor TJiP 1.2 2 trains R.V SR 3.3.1.4 NA NA 5reakers 3(a) 4 (a).
5(a) 2 trains C.V SR 3.3.1.4 NA NA
- 19. Reactor Trip 1.2 1 each U
SR 3.3.1.4 NA NA Breaker per RTB Undervultage and Shunt Trip 3(a)* 4(a). 5(a) 1 each C
SR 3.3.1.4 NA NA Mechanisms per RTB
- 20. Automatic Trip 10.2 2 trains C.V SR 3.3.1.5 NA NA Logic 3(a). 4 a)l 5(0) 2 trains C.V SR 3.3.1.5 NA NA (1) A channel is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel Is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(a) With Rod Control System capable of rod withdrawal. or one or sore rods not fully inserted.
(d)
Below the P-6 (Intermediate Range Neutron Flux) interlock.
(i)
Includin anr reactor trio b ss breakers that are racked 1n a n
dA tj) ow e!w tintermediate RanNeutron Flux) interlockW the logic inputs r
'Source Range Neutron FlyX LIIi-3.3-17 HBRSEP Unit No. 2 Amendment No. 176
__11 ESFAS INSTRUMENTATION 3.3.2 Table 3.3.2-1 (page I of 4)
Engineered Safety Feature Actuation System Instrumentation APPLICABLE HOOES OR IIOMINAL MTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE SETPOINT FUNCTION CONDITICOS CHAtNELS CONDITIONS REQUIREMIENTS VALUE (I)
- 1. Safety Injection
- a. Manual Initiation
- b. Automatic Actuation Logic and Actuation Relays
- c. Contarnment Pressure - High
- d. Pressurizer Pressure - Low 1.2.3.4 2
a SR 3.3.2.6 NA NA 1.2.3.4 2 trains 1.2.3.4 1.2.3"'
- e. Steam Line High Differential Pressure Between Steam Header and Steam Lines
- f. High Steam Flow in Two Steam Lines Coincident with Ta, - Low 3
3 3 per steam line 2 per steam line C
SR 3.3.2.2 SR 3.3.2.3 SR 3.3.2.5 E
SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 D
SIR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 D
SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 D
SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 O
SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 D
SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 D
SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 S 4.45 psig t 1709.89 psig 2 83.76 psig S 116.24 psig 4 psig 1715 psig 100 psig NA NA t b)
(b) 1.2
.3 tc)
(d)
Wb lb) 1.2
.3 I per loop 2 541.50 OF 543-F
- 9. High Steam Flow in Two Steam Lines Coincident with Steam Line Pressure - Low 1.2S.3b 2 per steam line 2 Sb).3(
1 per loop (c) td) k 614 psig psig (continued) tI) A channel is OPERABLE with an actual Trip Setpolnt value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to Its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(a) Above the Pressurizer Pressure interlock.
(b) Above the T"-.Low interlock.
Ce) Less than or equal to a function defined as AP corresponding to 41.58S full steam flow below 20S load. and AP Increasing linearly from 41.5a8 full steam flow at 20S load to 110.5 full stcam flow at IOOS load. and AP corresponding to 110.5S full steam flow above I1OS load.
Cd) A function defined as AP corresponding to 37.25S full steam flow between OS and 201 load and then a AP increasing linearly from 37.25S steam flow at 20S load to 1091 full steam flow at IOO load.
HBRSEP Unit No. 2 3.3-25 Amendment No. 196
United States Nuclear Regulatory Commission Attachment IV to Serial: RNP-RA/04-0067 6 Pages (including cover page)
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGES REGARDING REACTOR PROTECTION SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TABLES RETYPED TECHNICAL SPECIFICATIONS PAGES
RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 7)
Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES REQUIRED CONDITIONS SURVEILLANCE ALLOWABLE NOMINAL OR OTHER CHANNELS REQUIREMENTS VALUE TRIP SPECIFIED SETPOINT CONDITIONS (1)
- 1. Manual Reactor Trip 1.2 2
B SR 3.3.1.14 NA NA 3(a), 4(a), 5(a) 2 C
SR 3.3.1.14 NA NA
- 2. Power Range Neutron Flux
- a. High 1.2 4
D SR 3.3.1.1 110.93S 108%
SR 3.3.1.2 RTP RTP (2)
SR 3.3.1.7 SR 3.3.1.11
- b. Low 1 (b). 2 4
E SR 3.3.1.1 26.93%
SR 3.3.1.8 RTP 24S RTP SR 3.3.1.11
- 3. Intermediate Range 1(b) 2(c) 2 F.G SR 3.3.1.1 s 36.40%
25X RTP Neutron Flux SR 3.3.1.8 RTP SR 3.3.1.11 2(d) 2 H
SR 3.3.1.11 36.40k 25 RTP SR 3.3.1.8 RTP SR 3.3.1.11
- 4. Source Range Neutron SR 3.3.1.1 s 1.28 E5 1.0 E5 Flux 2(d) 2 I.J SR 3.3.1.8 cps cps SR 3.3.1.11 SR 3.3.1.1 1.28 ES 1.0 E5 3(a), 4 (a),
5(a) 2 J.K SR 3.3.1.7 cps cps SR 3.3.1.11 SR 3.3.1.1 N/A N/A 3(e) 4(e) 5(e) 1 L
SR 3.3.1.11 (continued)
(1)
A channel is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(2)
The Nominal Trip Setpoint is as stated unless reduced as required by one or more of the following requirements:
LCO 3.2.1 Required Action A.2.2; LCO 3.2.2 Required Action A.1.2.2: or LCO 3.7.1 Required Action B.2.
(a) With Rod Control System capable of rod withdrawal, or one or more rods not fully inserted.
(b) Below the P-10 (Power Range Neutron Flux) interlock.
(c) Above the P-6 (Intermediate Range Neutron Flux) interlock.
(d) Below the P-6 (Intermediate Range Neutron Flux) interlock.
(e) With the RTBs open. In this condition, source range Function does not provide reactor trip but does provide indication and alarm.
HBRSEP Unit No. 2 3.3-13 Amendment No.
RPS Instrumentation 3.3.1 FUNCTION APPLI(
OF SP Cot Table 3.3.1-1 (page 3 of 7)
Reactor Protection System Instrumentation
- ABLE MODES REQUIRED CONDITIONS SURVEILLANCE ALLOWABLE NOMINAL I OTHER CHANNELS REQUIREMENTS VALUE TRIP ECIFIED SETPOINT 4DITIONS (1)
- 9.
Reactor Coolant Flow - Low
- a. Single Loop
- b. Two Loops I(g)
I(h) 3 per loop 3 per loop N
SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 M
SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 2 93.45%
94.26%
2 93.45%
94.26%
- 10. Reactor Coolant Pump (RCP) Breaker Position
- a. Single Loop
- b.
Two Loops 1(g) 1(h) 1 per RCP 1 per RCP 0
SR 3.3.1.14 M
SR 3.3.1.14 NA NA NA NA
- 11. Undervoltage RCPs
- 12.
Underfrequency RCPs
- 13.
Water Level-Low Low 1(f) 1(f) 1.2 1 per bus 1 per bus 3 per SG M
SR 3.3.1.9 SR 3.3.1.10 M
SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 2 2959 V 3120 V 2 57.84 Hz 2 15.36%
58.2 Hz 16%
(continued)
(1) A channel is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(f) Above the P-7 (Low Power Reactor Trips Block) interlock.
(g) Above the P-8 (Power Range Neutron Flux) interlock.
(h) Above the P-7 (Low Power Reactor Trips Block) interlock and below the P-8 (Power Range Neutron Flux) interlock.
HBRSEP Unit No. 2 3.3-15 Amendment No.
RPS Instrumentation 3.3.1 Table 3.3.1.1 (page 4 of 7)
Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES REQUIRED CONDITIONS SURVEILLANCE ALLOWABLE NOMINAL OR OTHER CHANNELS REQUIREMENTS VALUE TRIP SPECIFIED SETPOINT CONDITIONS (1)
- 14.
SR 3.3.1.1 2 29.36%
30%
Level-Low SR 3.3.1.7 SR 3.3.1.10 Coincident with 1.2 2 per SG E
SR 3.3.1.1 7 01 E5 6.4 ES Steam SR 3.3.1.7 lbm/hr ibm/hr Flow/Feedwater Flow SR 3.3.1.10 Mismatch
- 15.
- a. Low Auto Stop 1(f) 3 P
SR 3.3.1.10 2 40.87 45 psig Oil Pressure SR 3.3.1.15 psig
- b. Turbine Stop Valve Closure 1(f) 2 P
SR 3.3.1.15 NA NA
- 16.
Safety 1.2 2 trains Q
SR 3.3.1.14 NA NA Injection (SI)
Input from Engineered Safety Feature Actuation System (ESFAS)
(continued)
(1) A channel is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(f) Above the P-7 (Low Power Reactor Trips Block) interlock.
I HBRSEP Unit No. 2 3.3-16 Amendment No.
RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 7)
Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES REQUIRED CONDITIONS SURVEILLANCE ALLOWABLE NOMINAL OR OTHER CHANNELS REQUIREMENTS VALUE TRIP SPECIFIED SETPOINT CONDITIONS (1)
- 17.
Reactor Protection System Interlocks
- a. Intermediate 2(d) 2 S
SR 3.3.1.11 2 9.34 1 E-10 Range Neutron SR 3.3.1.13 E-11 amp amp Flux. P-6
- b. Low Power 1
1 per T
SR 3.3.1.13 NA NA Reactor Trips train SR 3.3.1.14 Block. P-7
- c. Power Range 1
4 T
SR 3.3.1.11 s 42.94%
40% RTP Neutron Flux.
- d. Power Range 1.2 4
S SR 3.3.1.11 2 7.06%
10% RTP Neutron Flux.
- e. Turbine Impulse 1
2 T
SR 3.3.1.1 s 10.71%
10%
Pressure. P-7 SR 3.3.1.10 turbine turbine input SR 3.3.1.13 power power
- 18. Reactor Trip 1.2 2 trains R.V SR 3.3.1.4 NA NA Breakers(i) 3(a). 4(a), 5(a) 2 trains C.V SR 3.3.1.4 NA NA
- 19. Reactor Trip 1.2 1 each U
SR 3.3.1.4 NA NA Breaker per RTB Undervoltage and Shunt Trip 3(a) 4(a) 5(a) 1 each C
SR 3.3.1.4 NA NA Mechanisms per RTB
- 20. Automatic Trip
- 1. 2 2 trains Q.V SR 3.3.1.5 NA NA Logic 3(a) 4(a) 5(a) 2 trains CV SR 3.3.1.5 NA NA (1) A channel is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(a) With Rod Control System capable of rod withdrawal. or one or more rods not fully inserted.
(d) Below the P-6 (Intermediate Range Neutron Flux) interlock.
(i) Including any reactor trip bypass breakers that are racked in and closed for bypassing an RTB.
(j) Not used.
I I
HBRSEP Unit No. 2 3.3-17 Amendment No.
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 4)
Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR NOMINAL OTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE SETPOINT FUNCTION CONDITIONS ACTIONS CONDITIONS REQUIREMENTS VALUE (1)
- 1. Safety Injection
- a. Manual Initiation 1,2.3,4 2
B SR 3.3.2.6 NA NA
- b. Automatic 1.2.3.4 2 trains C
SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays
- c. Containment 1.2.3.4 3
- 4.45 psig 4 psig Pressure-High SR 3.3.2.4 SR 3.3.2.7
- d. Pressurizer 1,2,3(a) 3 D
SR 3.3.2.1 2 1709.89 psig 1715 psig Pressure-Low SR 3.3.2.4 SR 3.3.2.7
- e. Steam Line 1.2.3(a) 3 per D
SR 3.3.2.1 2 83.76 psig 100 psig High Differential steam SR 3.3.2.4
- 116.24 psig Pressure Between line SR 3.3.2.7 Steam Header and Steam Lines
- f. High Steam Flow in 1,2(b) 3(b) 2 per D
SR 3.3.2.1 (c)
(d)
Two Steam Lines steam SR 3.3.2.4 line SR 3.3.2.7 Coincident with 1,2 (b),
3(b) 1 per D
SR 3.3.2.1 2 541.50 *F 5431F Ta.- Low loop SR 3.3.2.4 SR 3.3.2.7
- g.
High Steam Flow in 1,2(b), 3(b) 2 per D
SR 3.3.2.1 (c)
(d)
Two Steam Lines steam SR 3.3.2.4 line SR 3.3.2.7 Coincident with 1,2 (b)* 3(b) 1 per D
SR 3.3.2.1 2 597.76 614 psig Steam Line loop SR 3.3.2.4 psig Pressure-Low SR 3.3.2.7 (continued)
(1) A channel is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(a) Above the Pressurizer Pressure interlock.
(b) Above the Tang-Low interlock.
(c) Less than or equal to a function defined as AP corresponding to 41.58X full steam flow below 20% load. and AP increasing linearly from 41.58% full steam flow at 20% load to 110.5% full steam flow at 100% load. and AP corresponding to 110.5% full steam flow above 100% load.
(d) A function defined as AP corresponding to 37.25% full steam flow between 0% and 20% load and then a AP increasing linearly from 37.25% steam flow at 20% load to 109% full steam flow at 100% load.
HBRSEP Unit No. 2 3.3-25 Amendment No.
United States Nuclear Regulatory Commission Attachment V to Serial: RNP-RA/04-0067 255 Pages (including cover page)
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGES REGARDING REACTOR PROTECTION SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TABLES SUPPORTING CALCULATIONS RNP-I/INST-1135, "Nuclear Instrumentation Intermediate Range Error Analysis" RNP-I/INST-1128, "RCS Flow Instrument Uncertainty and Scaling Calculation" RNP-I/INST-1041, "Feedwater Flow Loop Uncertainty and Scaling Calculation" RNP-I/INST-1043, "Main Steam Pressure Uncertainty and Scaling Calculation"