ML023240227
| ML023240227 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/12/2002 |
| From: | Keenan J Carolina Power & Light Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BSEP 02-0169, TSC-2002-09 | |
| Download: ML023240227 (47) | |
Text
John S Keenan Vice President Brunswick Nuclear Plant NOV 1 2 200Z SERIAL: BSEP 02-0169 TSC-2002-09 10 CFR 50.90 U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-62 Request For License Amendments Core Flow Operating Range Expansion
REFERENCES:
- 1. GE Nuclear Energy Report NEDC-33006P, "Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report,"
Revision 1, dated August 2002.
- 2. GE Nuclear Energy Report NEDC-33063, "Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 Maximum Extended Load Line Limit Analysis Plus," dated November 2002.
- 3. GE Nuclear Energy Report NEDC-33075P, "Detect And Suppress Solution - Confirmation Density Licensing Topical Report,"
Revision 2, dated November 2002.
- 4. GE Nuclear Energy Report NEDE-23785P-A, Vol. III, Supplement 1, Revision 1, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement 1, Additional Information for Upper Bound PCT Calculation," dated March 2002.
Ladies and Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Parts 50.90 and 2.101, Carolina Power & Light (CP&L) Company is requesting a revision to the Technical Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.
The proposed license amendments revise TSs, as necessary, to support an expansion of the core flow operating range (i.e., Maximum Extended Load Line Limit Analysis Plus (MELLLA+)).
The basis for the request was prepared following the guidelines contained in GE Nuclear Energy Report NEDC-33006P, "Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report," Revision 1, dated August 2002. Enclosure 2 contains GE PO Box 10429 Southport. NC 28461 T > 9104572496 F> 9104572803 SCP&L A Progress Energy Company
Document Control Desk BSEP 02-0169 / Page 2 Nuclear Energy Report NEDC-33063, "Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 Maximum Extended Load Line Limit Analysis Plus," dated November 2002 (i.e., M+SAR). The M+SAR is a summary of the results of the safety analyses performed for the BSEP core flow operating range expansion. The M+SAR contains information which GE Nuclear Energy considers to be proprietary. GE Nuclear Energy requests that the proprietary information in this report be withheld from public disclosure in accordance with 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1). An affidavit supporting this request is included in the M+SAR. The NRC may duplicate this submittal, including the M+SAR, for the purpose of internal review. Enclosure 3 contains a non-proprietary version of the M+SAR.
As part of MELLLA+ implementation for BSEP, CP&L will implement the Detect and Suppress Solution - Confirmation Density (DSS-CD) approach to automatically detect and suppress neutronic/thermal-hydraulic instabilities (THI). DSS-CD represents an evolutionary step from the Boiling Water Reactor Owners' Group (BWROG) Option III Reactor Stability Long-Term Solution, currently approved for use at BSEP. Section 2.4, "Stability," of the M+SAR discusses this change and verifies the applicability of GE Nuclear Energy Report NEDC-33075P, "Detect And Suppress Solution - Confirmation Density Licensing Topical Report," Revision 2, dated November 2002, to BSEP Units 1 and 2. Upon issuance of MELLLA+ for BSEP, the Backup Stability Protection (BSP) described in NEDC-33075P will be the preferred alternate method to detect and suppress THI oscillations as allowed by TS 3.3.1.1 Action I. CP&L intends to operate for up to one cycle, for each unit, with the new DSS-CD trip function bypassed. The will allow sufficient time to collect operational data necessary to validate trip setpoints and, thereby, avoid unnecessary reactor scrams. The proposed license amendments revise TSs, as necessary, to support implementation of DSS-CD at BSEP.
The M+SAR also documents the applicability of GE Nuclear Energy Report NEDE-23785P-A, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement 1, Additional Information for Upper Bound PCT Calculation," to BSEP Units 1 and 2. As a result, the 1600 degree F, upper bound peak clad temperature NRC limit is removed. Further discussion of this subject can be found in Section 4.3, "Emergency Core Cooling System Performance," of the M+SAR.
CP&L has evaluated the proposed change in accordance with 10 CFR 50.9 l(a)(1), using the criteria in 10 CFR 50.92(c), and determined that this change involves no significant hazards considerations.
CP&L plans to implement MELLLA+ during the Unit 1 Cycle 15 Refueling Outage (i.e.,
B1 15R1, currently scheduled to begin in February 2004) and the Unit 2 Cycle 17 Refueling Outage (i.e., B217R1, currently scheduled to begin in March 2005). However, to support required reload core analyses for Unit 1, CP&L requests that these amendments be issued by October 1, 2003.
Document Control Desk BSEP 02-0169 / Page 3 CP&L requests that the amendments, once approved, be issued effective immediately, to be implemented prior to startup from the B115R1 refueling outage for Unit 1 and prior to the startup from the B217R1 refueling outage for Unit 2.
In accordance with 10 CFR 50.9 l(b), CP&L is providing the State of North Carolina a copy of the proposed license amendments.
Please refer any questions regarding this submittal to Mr. Edward T. O'Neil, Manager - Support Services, at (910) 457-3512.
Sincerely, J S. Keetan MAT/mat
Enclosures:
- 1. Evaluation of Proposed License Amendment Request
- 2. NEDC-33063, "Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 Maximum Extended Load Line Limit Analysis Plus," dated November 2002 Proprietary
- 3. NEDO-33063, "Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 Maximum Extended Load Line Limit Analysis Plus," dated November 2002 Non-Proprietary
- 4. Marked-up Technical Specification Pages - Unit 1
- 5. Marked-up Technical Specification Pages - Unit 2
- 6. Marked-up Technical Specification Bases Pages - Unit 1 (For Information Only)
- 7. List of Regulatory Commitments John S. Keenan, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, and agents of Carolina Power &
Light Company.
Notary (Seal)
My commission expires: a4 U-14 2.,
% 0"4-
Document Control Desk BSEP 02-0169 / Page 4 cc (with enclosures except as noted):
U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTIN: Mr. Theodore A. Easlick, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission ATTN: Leonard N. Olshan (Mail Stop OWFN 8H12) 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission ATTN: Allen G. Howe (Mail Stop OWFN 8G9) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford (w/o Enclosure 2)
Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Ms. Beverly 0. Hall, Section Chief (w/o Enclosure 2)
Radiation Protection Section, Division of Radiation Protection North Carolina Department of Environment and Natural Resources 3825 Barrett Drive Raleigh, NC 27609-7221
BSEP 02-0169 Page 1 of 7 Evaluation of Proposed License Amendment Request
Subject:
Request For License Amendments Core Flow Operating Range Expansion 1.0 Description This letter is a request to amend Operating Licenses DPR-71 and DPR-62 for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.
The proposed license amendments revise Technical Specifications (TSs), as necessary, to (1) support an expansion of the core flow operating range and (2) implement the Detect and Suppress Solution - Confirmation Density (DSS-CD) approach to automatically detect and suppress neutronic/thermal-hydraulic instabilities (THI). These changes will support operation of Units 1 and 2 at 2923 megawatts thermal (MWt) with core flow as low as 85 percent of rated core flow (i.e., Maximum Extended Load Line Limit Analysis Plus (MELLLA+)).
2.0 Proposed Change To support an expansion of the core flow operating range for Units 1 and 2, Carolina Power &
Light (CP&L) Company requests the following TS changes.
PrpoedTchnica1 Spefctio Cange ML A+~
I....
. o TS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation" Table 3.3.1.1-1, Function 2.b
< 0 55W + 62.6% RTP and
< 0 61W + 65.2% RTP and (Average Power Range Monitors,
< 117.1% RTP
< 117.1% RTP Simulated Thermal Power - High)
Allowable Value Table 3.3.1.1-1, Note b
< [0.55 (W - AW) + 62 6% RTP]
< [0.61 (W - AW) + 65.2% RTP]
when reset for single loop operation when reset for single loop operation per LCO 3.4 1, "Recirculation Loops per LCO 3.4.1, "Recirculation Loops Operating." The Value of AW is Operating." The Value of AW is defined in plant procedures.
defined in plant procedures.
Justification:
M+SAR Section 5.3.1
BSEP 02-0169 Page 2 of 7 S...................
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~.4PpOe TS 3.4.1, "Recirculation Loops Operating" LCO 3.4.1 Two recirculation loops with Two recirculation loops with matched recirculation pump speeds matched recirculation pump speeds shall be m operation, shall be m operation, OR OR One recirculation loop may be in One recirculation loop may be in operation provided the following operation provided the plant is not limits are applied when the operating in the MELLLA+ region associated LCO is applicable:
defined in the COLR and the following limits are applied when the associated LCO is applicable Justification:
M+SAR Section 3.6 TS 3.4.3, "Safety/Relief Valves (SRVs)"
LCO 3.4.3 The safety function of 10 SRVs shall The safety function of 11 SRVs shall be OPERABLE.
be OPERABLE.
Actions A. One or more required SRVs A. One SRV inoperable.
A. 1 Exit the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> A. 1 Be in MODE 3 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> MELLLA+ operating AND region defined in the A.2 Be in MODE 4 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> COLR B. Two or more SRVs inoperable.
OR Required Action and associated Completion Time of Condition A not met.
B.1 Bein MODE 3 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND B.2 Be in MODE 4 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SR 3.4.3.1 Verify the safety function lift Verify the safety function lift setpoints of the required 10 SRVs setpoints of the 11 SRVs are as are as follows:
follows:
Justification:
M+SAR Section 9.3.1
BSEP 02-0169 Page 3 of 7 To support implementation of the DSS-CD approach to automatically detect and suppress THI, CP&L requests the following TS changes.
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.chna c.iato s-S..t TS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation" SR 3.3.1.1.19 Verify OPRM is not bypassed when Deleted.
APRM Simulated Thermal Power is
>25% and recirculation drive flow is Justification:
_<60%
M+SAR Section 2.4 and NEDC-33075P (Reference 4)
TS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation" Table 3.3.1.1 -1, Function 2.f (Average Power Range Monitors, OPRM - Upscale)
Surveillance Requirements SR 3.3.1.1.19 Delete Note (d)
See COLR for OPRM period based Delete detection algorithm (PBDA) setpoint limits.
Justification:
M+SAR Section 2.4 and NEDC-33075P (Reference 4) iiiiiiiliiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiii i ~ii
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TS 5.6.5 Core Operating Limits Report (COLR)
- a. Core operating limits shall be No current requirement.
- 4. The Maximum Extended Load established prior to each reload Line Limit Analysis Plus cycle, or prior to any remaining (MELLLA+) operating region for portion of a reload cycle, and Specifications 3.4.1 and 3.4.3.
shall be documented in the COLR for the following:
Justification:
Consistency with M+SAR Sections 3.6 and 9.3.1.
BSEP 02-0169 Page 4 of 7 1~reo~ T~hnk Sp~c~ito~i....g.....SS&
Spcfiti:.
I.at 1'dtin RequiDmeS
.~pe
...... ni TS 5.6.5 Core Operating Limits Report (COLR)
- a. Core operating limits shall be
- 3. The period based detection Delete established prior to each algorithm (PBDA) setpoint for reload cycle, or prior to any Function 2.f, Oscillation Power Justification:
remaining portion of a reload Range Monitor (OPRM) Upscale, M+SAR Section 2.4 and cycle, and shall be for Specification 3.3.1.1; and NEDC-33075P (Reference 4) documented in the COLR for the following:
In summary, the proposed changes:
(1) delete SR 3.3.1.1.19, reference to SR 3.3.1.1.19 in TS Table 3.3.1.1 -1, and Note d of TS Table 3.3.1.1-1, to account for implementation of the DSS-CD approach to THI, (2) modify the Average Power Range Monitor (APRM) scram flow-biased setpoint allowable values for two and single recirculation loop operation, in TS Table TS 3.3.1.1-1, to accommodate the expanded operating range, (3) revise TS 3.4.1 to restrict single loop operation to areas outside of the MELLLA+
operating region, (4) revise the SRV operability and surveillance requirements, of TS 3.4.3, to reflect the need to maintain 11 operable SRVs while in the MELLLA+ operating region, (5) delete the existing TS 5.6.5.a.3 which requires the COLR to include the period based detection algorithm (PBDA) made obsolete with implementation of the DSS-CD approach to TIl, and (6) add a new requirement (i.e., TS 5.6.5.a.4) to established the MELLLA+ operating region in the COLR.
These changes support the MELLLA+ expansion of the core flow operating range and implement the DSS-CD approach to automatically detect and suppress THI for BSEP Units 1 and 2 CP&L will make supporting changes to the TS Bases in accordance with TS 5.5.10, "Technical Specifications (TS) Bases Control Program." Enclosure 5 provides marked-up TS Bases pages for Unit 1. These pages are being submitted for information only and do not require issuance by the NRC.
BSEP 02-0169 Page 5 of 7 3.0
Background
On May 31, 2002, the NRC issued License Amendments 222 and 247 to the BSEP Unit 1 and 2 TSs (i.e., Reference 1). The amendments allowed an increase in the maximum power level of each unit from 2558 MWt to 2923 MWt. In support of the power uprate, the power/flow operating map (i.e., a licensing region which defines boundaries, inside of which operation of the plant has been analyzed and demonstrated to meet all applicable fuel and system design criteria) was revised. The revised power/flow operating map provides only a narrow operating band, with respect to core flow (i.e., approximately 99 percent to 102.2 percent core flow), when operating at 2923 MWt. This narrow operating band unduly restricts plant operation. In order to restore operating flexibility, CP&L is requesting an expansion of the core flow operating range.
4.0 Technical Analysis As stated above, CP&L is requesting a revision to power/flow operating map to restore operating flexibility at the CLTP of 2923 MWt. This is accomplished by permitting operation at the CLTP with core flow as low as 85 percent of rated core flow versus approximately 99 percent as limited by the existing power/flow operating map. The expanded operating range is identified as MELLLA+. As a result of this change there will be:
"* No increase in the current maximum normal operating reactor dome pressure,
"* No increase in core power,
"* No increase in the maximum licensed core flow,
"* No change to source term methodology,
"* No new fuel product line,
"* No change in fuel cycle length, and
"* No additions to currently licensed operational enhancements.
The basis for the request was prepared following the guidelines contained in GE Nuclear Energy Report NEDC-33006P, "Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report," Revision 1, dated August 2002 (i.e., Reference 2). Enclosure 2 contains GE Nuclear Energy Report NEDC-33063, "Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 Maximum Extended Load Line Limit Analysis Plus," dated November 2002 (i.e.,
Reference 3, referred to as the M+SAR). The M+SAR is a summary of the results of the safety analyses performed for the BSEP core flow operating range expansion. In preparation of the M+SAR, evaluations of the reactor, engineered safety features, power conversion, emergency power, support systems, environmental issues, design basis accidents, and previous licensing evaluations were performed. All safety aspects of the plant, affected by MIELLLA+, were evaluated using previously NRC-approved or industry-accepted analytical methods.
As part of MELLLA+ implementation for BSEP, CP&L will also implement the DSS-CD approach to automatically detect and suppress THI. DSS-CD represents an evolutionary step
BSEP 02-0169 Page 6 of 7 from the BWROG Option III Reactor Stability Long-Term Solution, currently approved for use at BSEP. The DSS-CD approach uses the same hardware design as Option III; it introduces an enhanced detection algorithm (i.e., the Confirmation Density Algorithm (CDA)) to detect the inception of power oscillations and generate an earlier power suppression trip signal based on successive period confirmation recognition. The existing Option III algorithms are retained to provide defense-in-depth protection. The DSS-CD approach will continue to provide reliable, automatic detection and suppression of stability related power oscillations and provide protection against violation of the Safety Limit Minimum Critical Power Ratio (SLMCPR) for anticipated oscillations.
Section 2.4, "Stability," of the M+SAR discusses this change and verifies the applicability of GE Nuclear Energy Report NEDC-33075P, "Detect And Suppress Solution - Confirmation Density Licensing Topical Report," Revision 2, dated November 2002 (i.e., Reference 4), to BSEP Units 1 and 2.
5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration Section 11.3.3, "Assessment of 10 CFR 50.92 Criteria," of the M+SAR evaluates whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment." Based on this evaluation, CP&L concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
5.2 Applicable Regulatory Requirements/Criteria Implementation of MELLLA+ does not require (1) an increase in the current maximum normal operating reactor dome pressure, (2) an increase in core power, (3) an increase in the maximum licensed core flow, (4) a change to source term methodology, (5) a new fuel product line, or (6) a change in fuel cycle length. As such, the impact on plant operation is minimal and, as demonstrated in the M+SAR, the MELLLA+ range expansion can be accomplished without exceeding any existing regulatory limits or design allowable limits applicable to BSEP.
As part of MELLLA+ implementation for BSEP, CP&L will also implement the DSS-CD approach to automatically detect and suppress THI. Since the DSS-CD approach will continue to provide reliable, automatic detection and suppression of stability related power oscillations and provide protection against violation of the SLMCPR for anticipated oscillations, compliance with General Design Crieria (GDC) 10 and 12 of 10 CFR 50, Appendix A is maintained.
This request is not being submitted as a risk informed licensing action, as defined by Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated July 1998 (i.e., Reference 5).
However, it was evaluated from a risk perspective and, as demonstrated in Section 10.5,
BSEP 02-0169 Page 7 of 7 "Individual Plant Evaluation," of the M+SAR, the net increase in core damage frequency (CDF) and large early release frequency (LERF) are not significant.
6.0 Environmental Considerations A review has determined that the proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.
7.0 References
- 1.
Letter from the U. S. Nuclear Regulatory Commission to Mr. John S. Keenan, "Issuance Of Amendment Re: Extended Power Uprate," dated May 31, 2002. (ML021430551).
- 2.
GE Nuclear Energy Report NEDC-33006P, "Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report," Revision 1, dated August 2002.
- 3.
GE Nuclear Energy Report NEDC-33063, "Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 Maximum Extended Load Line Limit Analysis Plus," dated November 2002.
- 4.
GE Nuclear Energy Report NEDC-33075P, "Detect And Suppress Solution - Confirmation Density Licensing Topical Report," Revision 2, dated November 2002.
- 5.
Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated July 1998.
Precedents BSEP is the lead plant for MELLLA+ and DSS-CD implementation.
BSEP 02-0169 Marked-up Technical Specification Pages - Unit 1
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE SR 3.3.1.1.19 V.erify OPRM is Rot bypassed when APRM 1
muted Therinal Power is 7Ž. 25% aft4
.rcceulatien drivc flew 4s -5 699-FREQUENCY t 4 meniths Brunswick Unit 1 I
3.3-8 Amendment No.--*,ý
RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)
Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 1.
- a.
Neutron Flux--High 2
5 (a)
- b.
Inop 2
5 (a)
- 2.
Average Power Range Monitors
- a.
Neutron Flux--High (Setdown)
- b.
Simulated Thermal Power -High 2
1 3
3 3
3 3 (c)
G SR SR SR SR SR SR SR H
SR SR SR SR SR G
SR SR SR H
SR SR SR G
SR SR SR SR SR SR 3.3.1.1.2 3.3.1.1.4 3.3.1.1.5 3.3.1.1.6 3.3.1.1.7 3.3.1.1.13 3.3.1.1.15 3.3.1.1.2 3.3.1.1.4 3.3.1.1.5 3.3.1.1.13 3.3.1.1.15 3.3.1.1.4 3.3.1.1.5 3.3.1.1.15 3.3.1.1.4 3.3.1.1.5 3.3.1.1.15 3.3.1.1.2 3.3.1.1.5 3.3.1.1.7 3.3.1.1.8 3.3.1.1.11 3.3.1.1.13 S120/125 divisions of full scale 5 120/125 divisions of full scale NA NA S22.7% RTP (a)
With any control rod withdrawn from a core cell containing one or more fuel assemblies.
(b) fZ.kj, when reset for single loop operation per LCO 3.4.1, "Reci ing." The value of AW is defined in plant procedures.
rculation Loops (c)
Each APRM channel provides inputs to both trip systems.
Brunswick Unit 1 I
I 3.3-9 Amendment No. HiJl,
RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)
Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 2.
Average Power Range Monitors (continued)
- c.
Neutron Flux -- High
- d.
Inop
- e.
2-Out-Of-4 Voter
- f.
OPRM Upscale
- 3.
Reactor Vessel Steam Dome Pressure -High
- 4.
Reactor Vessel Water Level -- Low Level 1
- 5.
Main Steam Isolation Valve -- Cosure
- 6.
Drywett Pressure--High 1
3 (c) 1,2 3 (c) 1,2 2
S20% RTP 3 (c) 1,2 1,2 1
1,2 2
2 8
2 F
SR 3.3.1.1.2 SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 G
SR 3.3.1.1.5 SR 3.3.1.1.11 G
SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.17 SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 G
_SR 3.3.1.1.1R SR 3.3.1.1.5 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17 SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17 SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 S118.7X RTP NA NA NAi
- 1077 psig
? 153 inches 4 lOX closed S 1.8 psig (continued)
(c)
Each APRM channel provides inputs to both trip systems.
cl G-: CQGl 48. OPflM pepiv based fmerik eal h
.:h.(rDD9.) setAr~i t 1i q.
Brunswick Unit 1 I
I Amendment No.
3.3-10
Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO
3.4.1 APPLICABILITY
Two recirculation loops with matched recirculation pump speeds shall be in operation, OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:
LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR),"
single loop operation limits specified in the COLR;
- b.
LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; and
- c.
LCO 3.3.1.1, "Reactor Protection System (RPS)
Instrumentation," Function 2.b (Average Power Range Monitors Simulated Thermal Power-High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.
MODES 1 and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirements of the A.1 Satisfy the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> LCO not met.
requirements of the LCO.
(continued)
Brunswick Unit I I
Amendment No.".4-'l 3.4-1
SRVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (SRVs)
LCO
3.4.3 APPLICABILITY
ACTIONS CONDIT A.
One SRV)inoperabe.
The safety function of SRVs shall be OPERABLE.
6xii MODES 1, 2-and 3.
OL reIL do Liz, rcý UIRED ACTION t
7 A n P0- in MQflF 4 COMPLETION TIME 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify t afet function lift setpoints In accordance of the SRVs are as follows:
with the Inservice Number of Setpoint Testing Program SRVs (psig) 4 1130 +/- 33.9 4
1140 +/- 34.2 3
1150 +/- 34.5
- ., 1 o
,R'&
r13.
1 Be -, r~ob_
k (continued) as Brunswick Unit 1 3.4-5 Amendment No. f0 B
19-1*"
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)
- a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
- 1.
The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1;
- 2.
The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.2; or Function 2.f, Oscillatio-no.,cr n.
ng f-f-oI
)
)Ups.alc, for-
.pccifatien 3.3.4.1. and The Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions for Specification 332l
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1.
NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (latest approved version).
C; The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
(continued) f~a-gimum Lh-L- ;-t /Rn~
L 5P
- q. v (AEL~4 Z
L
~OL;
-o Brunswick Unit I Amendment No.-*
5.0-20
BSEP 02-0169 Marked-up Technical Specification Pages - Unit 2
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE SR 3.3.1.1.19 Y.erify OPWI1 is no~t b Simulated Therm-al Po Pccivcuhation driye yvase whenI, APR!D wer is : 25% an#
d flow is !5 O%.
FREQUENCY 4
t 4 month I-Brunswick Unit 2 Amendment No.-Z*
3.3-8
If RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)
Reactor Protection System Instrumentation APPLICABLE CONDITIONS MOVES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOUABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 1.
- a.
Neutron Ftux-High 2
3 G
- s 120/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.5 full scale SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.15 5 (a)
_< 120/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.5 full scale SR 3.3.1.1.13 SR 3.3.1.1.15
- b.
Inop 2
3 G
SR 3.3.1.1.4 NA SR 3.3.1.1.5 SR 3.3.1.1.15 5 (a) 3 H
SR 3.3.1.1.4 NA SR 3.3.1.1.5 SR 3.3.1.1.15
- 2.
Average Power Range Monitors
- a.
Neutron Flux--High 2
3 (c)
< 22.7% RTP (Setdown)
SR 3.3.1.1.5 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 33111 SR 3.3.1.1.13
- b.
Simulated Thermal 1
3(c)
F S 3.3.1.1.
b)
Power -High SR 3.3.1.1.
- 117.1% RTP S.3.3.1.1.11 SRg 3.3.1.1.13 S...
SR 3.3.1.1.18
<0.41 VI+ 4,5',a% rT?
(continued)
(a)
With any control rod withdrawn from a core cell containing one or more fuel assemblies.
(b) rwhen reset for single loop operation per LCO 3.4.1, "Recirculation Loops Ov perating.n The 1alue4 of AW is defined in plant procedures.
(c ahAPPM charnel provides inputs to both trip systems.
Brunswick Unit 2 I
3.3-9 Amendment No._-Zý
RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)
Reactor Protection System Instrumentation APPLICABLE CONDITIONS MOOES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 2.
Average Power Range Monitors (continued)
- c.
Neutron Flux -High
- d.
Inop
- e.
2-Out-Of-4 Voter
- f.
OPRM Upscale
- 3.
Reactor Vessel Steam Dome Pressure -Nigh
- 4.
Reactor Vessel Water Level -Low Level 1
- 5.
Main Steam Isolation Valve -Closure
- 6.
Drywett Pressure -High 1
3 (c) 1,2 3 (c) 1,2 2
S20% RTP 3 (c) 1,2 1,2 1
1,2 2
2 8
2 F
SR 3.3.1.1.2 SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 G
SR 3.3.1.1.5 SR 3.3.1.1.11 G
SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.17 SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18 G
SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17 G
SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17 F
SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.17 G
SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 S118.7% RTP NA NA NA'E=
- 1077 psig S153 inches S10% closed
- 1.8 psig (continued)
(c)
Each APRM charnel provides inputs to both trip systems.
(~d~~
C!R$or. DrPR? paried based dzet~ztm elszr4thm (PBD:.) ge.tperA limi Brunswick Unit 2 I
I 3.3-10 Amendment No.
- 34/I
Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched recirculation pump speeds shall be in operation, OR One recirculation loop may be in operation providedhe following limits are applied when the associated LCO is applicable:
- a.
COLR; "AVERAGE PLANAR LINEAR HEAT GENERATION RATE single loop operation limits specified in the
- b.
LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"
single loop operation limits specified in the COLR; and
- c.
LCO 3.3.1.1, "Reactor Protection System (RPS)
Instrumentation," Function 2.b (Average Power Range Monitors Simulated Thermal Power-High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.
APPLICABILITY:
MODES I and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirements of the A.1 Satisfy the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> LCO not met.
requirements of the LCO.
(continued)
Amendment No.-2ý1 Brunswick Unit 2 i
3.4-1
SRVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (SRVs)
LCO 3.4.3 The safety function offlSRVs shall be OPERABLE.
SRVS inoperabl e.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the afet function lift setpoints In accordance of the(.*cquircd 4 SRVs are as follows:
with the Inservice Number of Setpoint Testing Program SRVs (psig) 4 1130 +/- 33.9 4
1140 +/- 34.2 3
1150 +/- 34.5 0-(continued)
Reydrec OC iok~y40 13,&%
ur-St~t)q ass 0 Ltci C-4r&
V~ei 0Jo V... D I!
C.ca isi~ ft
)mU Me-t.
Brunswick Unit 2 3.4-5 Amendment No.
e-
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
CORE OPERATING LIMITS REPORT (COLR)
- a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
- 1.
The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1;
- 2.
The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.2;
..h.
peri.d based dzt..tin rlgecrit*h (PB..**
cetpoint-fo5Fntin i.,
Ozcillatien Pce.;c Range The Allowable Values and power range setpoints Rod Block Monitor Upscale Functions for Specification 3.3.2.1*-r*fEa
)
for
- b. The analytical methods used t operating limits shall be thc approved by the NRC, specifi(
the following documents:
- 1.
NEDE-24011-P-A, "General Application for Reactor version).
C. The core operating limits sh*
all applicable limits (e.g.,
limits, core thermal hydrauli Cooling Systems (ECCS) limit SDM, transient analysis limil limits) of the safety analys
- d.
The COLR, including any midc, supplements, shall be providi reload cycle to the NRC.
- q. Tke.
Ecf~~gtro(eoL L.,ca ot L; -t Pjv* (r6.1)*
oe-at rf rS2&Wý Brunswick Unit 2
- o determine the core se previously reviewed and tally those described in Electric Standard Fuel" (latest approved ill be determined such that fuel thermal mechanical ic limits, Emergency Core
- , nuclear limits such as ts, and accident analysis is are met.
ycle revisions or ed upon issuance for each (continued) 4Aen mpec No.
3 Amendment No.
-4z I
5.6.5 0---*GD I
I 5.0-19
BSEP 02-0169 Marked-up Technical Specification Bases Pages - Unit 1 (For Information Only)
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, Average Power Range Monitor (APRM)
(continued)
LCO, and channel is assigned to one, two or four OPRM "cells,"
APPLICABILITY forming a total of 24 separate OPRM cells per APRM channel, each with either three or four detectors.
LPRMs near the edge of the core are-assigned to either one or two OPRM cell s.
9ina PMeanlms ha' R t least tc O ER.[ LPflts fer thez EPRI! Upscale r.unt..*l L2.
i.lf ti be V P LE OIPR B,
,h LE ( n "e, IFR.
22 )L.
2.a.
Average Power Range Monitor Neutron Flux-High 0 *(Setdown) 0-For operation at low power (i.e., MODE 2), the Average Power Range Monitor Neutron Flux-High (Setdown) Function is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range.
For most operation at low power levels, the Average Power Range Monitor Neutron Flux-High (Setdown)
U, 0 0 Function will provide a secondary scram to the Intermediate 4-C06 Range Monitor Nbutron Flux-High Function because of the "relative setpoints.
With the IRMs at Range 9 or 10, it is possible that the Average Power Range Monitor Neutron
_WU -I Flux-High (Setdown) Function will provide the primary trip signal for a core-wide increase in power.
4p No specific safety analyses take direct credit for the CZ Average Power Range Monitor Neutron Flux-High (Setdown) ac Function.
However, this Function is credited in S41calculations used to eliminate the need to perform the 0
-spatial analysis required for the Intermediate Range Monitor ONeutron Flux-High Function (Ref. 6).
In addition, the Z 16 Average Power Range Monitor Neutron Flux-High (Setdown)
Function indirectly ensures that before the reactor mode
'~
0switch is placed in the run position, reactor power does not
~,..J
~exceed 23% RTP (SL 2.1.1.1) when operating at low reactor pressure and low core flow.
Therefore, it indirectly prevents fuel damage during significant reactivity increases 0-.C' with THERMAL POWER < 23% RTP.
0' 0
The Allowable Value is based on preventing significant S-4.
increases in power when THERMAL POWER is < 23% RTP.
S
- 4)
The Average Power Range Monitor Neutron Flux-High (Setdown)
~ ~.z' "Z I,7 Functinn mtit hbe NPFRARI F drynn Mnnf 9,9 n r-nninl i-nic may be withdrawn since the potential for criticality exists.
(continued)
Revision No.9-I-
Brunswick Unit I B 3.3-8
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a.
Average Power Range Monitor Neutron Flux-High SAFETY ANALYSES, (Setdown)
(continued)
LCO, and APPLICABILITY In MODE 1, the Average Power Range Monitor Simulated Thermal Power-High and Neutron Flux-High Functions provide protection against reactivity transients and the RWM and Rod Block Monitor protect against control rod withdrawal error events.
2.b.
Average Power Range Monitor Simulated Thermal Power-High The Average Power Range Monitor Simulated Thermal Power-High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant.
The APRM neutron flux is electronically filtered with a time constant, nominally 6 seconds, representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor.
The trip level is varied as a function of rated recirculation drive flow (W) in percent and is clamped at an upper limit that is always lower than the Average Power Range Monitor Neutron Flux-High Function Allowable Value.
The Average Power Range Monitor Simulated Thermal Power-High Function provides a general definition of the licensed core power/core flow operating domain.
A note is included, applicable when the plant is in single recirculation loop operation per LCO 3.4.1, which requires reducing byAW the flow value used in the Allowable Value equation.
The value ofAW is defined in plant procedures.
The value of i*W is established to adjust the SLO limit down Sin power approximaley RTP to reflect the difference between the analyzed limiis for two-recirculation loop operation (TLO) and SLO.
The adjustment maintains the SLO limits at approximately the same absolute thermal power evel as was established prior to extended power uprate.
The 8.5 RTP has been converted to an equivalent "Z1W" value for convenience of representation and to reflect the way the adjustment is actually made in the APRM equipment.
In addition to this adjustment, the actual MW value entered into the equipment includes an allowance for additional flow measurement uncertainties that may occur in-SLO.
The (continued)
Brunswick Unit I Revision No. T l
B 3.3-9
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.e.
2-Out-Of-4 Voter (continued)
SAFETY ANALYSES,
- LCO, and independently voted sets of Functions, each of which is APPLICABILITY redundant (four total outputs).
The analysis in Reference 15 took credit for this redundancy in the justification of the 12-hour Completion Time for Condition A, so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable.
The voter Function 2.e does not need to be declared inoperable due to any failure affecting only the APRM Interface hardware portion of the Two-Out-Of-Four Logic Module.
There is no Allowable Value for this Function.
2.f.
Oscillation Power Range Monitor (OPRM)
Upscale The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations.
detection lgorithm. Th.....
e rern algorithms pr..ide Ir,
~
~
~
dfes in depet~g theran-hdraddtil 4nst~ablty eate. neai r ee11_
,PEBILIY fa Spsi........ purposes is based se_-rilonl os~ains the period based dcetio tn algorithm.
h The OPRM Upscale Function receives input signals from the LPRMs within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms.
Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing S37 a trip signal before the MCPR SL is exceeded.
Three of the four channels are required to be OPERABLE.
The OPRM Upscale trip iscautoma inplly enabled (bypass removed) when THERMAL POW his c aRIP, as indicated by the APRM Simulated Thermal Power, and reactor core flow is de n trated flow, as indicated by APRM measured (continued)
Brunswick Unit 1 B 3.3-13 Revision No.4-1
Insert 1 Reference 20 describes the primary algorithm used in the OPRM for detecting thermal-hydraulic instability related neutron flux oscillations: the confirmation density algorithm. References 17, 18, and 19 describe three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All four algorithms are implemented in the OPRM Upscale Functions, but the safety analysis only takes credit for the confirmation density algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the confirmation density algorithm.
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f.
Oscillation Power Range Monitor (OPRM)
Upscale SAFETY ANALYSES, (continued)
,LCO, and APPLICABILITY recirculation drive flow.
This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur.
See Reference for additional discussion of OPRM Upscale trip enable re ion limits.
0 T
ee ----
MyRH__--e R6
.e L) tab.i.h -.
d b.c t
3 aluc in
-f-e.
2f
-f4
~y upratcd poer to srcpend to 30% of erizglinl plant RTFP.
T-his ___1... is nect reurdb Rcfcrcrnz 21, but has been.
done forcoertim These ý'etpoints, which are sometimes referred to as the "auto-bypass" setpoints, establish the boundaries of the OPRM Upscale trip enabled region.
The APRM Simulated Thermal Power auto-enable setpoint has 1% deadband while the drive flow setpoint has a 2% deadband.
The deadband for these setpoints (a& established so that it increased the enabled region.
The OPRM Upscale Function is required to be OPERABLE when the plant is at ;_20% RPT.
The 20% RTP level is selected to provide margin in the unlikely event that a reactor power increase transient occurring while the plant is operating be o RTP causes a power increase to or beyond the APR Simulated Thermal Power OPRM Upscale trip auto-enable setpoint without operator action.
This OPERABILITY requirement assures that the OPRM Upscale trip auto-enable function will be OPERABLE when required.
An OPR?-1 Upseale trip is issued-frm an ARAM channel when the rce..
period base detection algorithm in that channel detect:
oge:
in on flux, irdieetJ by the C_ DW11-oeebined signal: of the LPRM dctectors; in a cell, with period conf--irmation: an.
d el.aote:
cel amplitude exceeding specified setpeincts.
One or in a channel e
Ups-c-a eFunction 2.f in one APR14 channel in a (continued)
Brunswick Unit 1 R&PJ Revision No. -21Vl R 3.3-14
Insert 2 The 23% RTP lower boundary of the enabled region was established at the lowest value at which APRM calibration to core thermal power is performed.
Insert 3 An OPRM Upscale trip is issued from an APRM channel when the confirmation density algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmation confirmed in a number of OPRM cells exceeding a preset value. An OPRM Upscale trip is also issued from the channel if any of the other three OPRM algorithms detect oscillatory changes in the neutron flux for one or more cells in that channel.
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f.
Oscil SAFETY ANALYSES, (continued)
LCO, and
(
lation Power Range Monitor (OPRM)
Upscale "tripped" state, if necessary to satisfy a Required Action, the APRM equipment is conservatively designed to force an OPRM Upscale trip output from the APRM channel if an APRM Inop condition occurs, such as when the APRM chassis keylock switch is placed in the Inop position.)
I r
!i- "
^,..W Annnu IIIý+^
-L nUS+
a IU 1
u adjustment parameters-(a) OPRM trip auto enable sctpoints for simfulate hra oc (SIP) (25%) and drivc ls (69%); (b) period based detection algorithm (PBDA) confirmation count and amnplitudc setpoints; (e) pcricd base-'
dctcction algorithmf tuning' paccs and Ed) growth rat algorithm (CFU\\)
and amplitude-based algat-ith1 m (ABAt)
The first set, the OPRII auto enable rcgion setpeints, as sctpoints with no additional margins added. 46~ sctings, 25% APR11 Simulated Thermal Power and 6N%
drilsc flow,' a rc defi4ned (limit values) in an~d cnf~irmedbyZ
.111.
The second set, the GPR1M PBDA tlri --
epo ints, arc Referernza 23, and are deeumcrnted in the COLR. There ar.n
-allowable values for-these setpoints. The third set, the
'OPRI-PBDA "tuning" par'ameters, ere estftblished, adjusted, and ABA setpoints, in accordance with References 15 and 16&,
aro established as nominal Yalues enly, and eenratlled by
-plant pro.edures.
- 3.
Reactor Vessel Steam Dome Pressure-High An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion.
This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB.
The Reactor Vessel Steam Dome Pressure-High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power.
For the (continued)
Revision No.-'aI APPLICABILITY i?.rlckcL:. w,7%
- ser-I L(
I Brunswick Unit I B 3.3-15
Insert 4 There are four "sets" of OPRM related setpoints or adjustment parameters: (a) OPRM trip auto enable setpoints for STP (23%) and drive flow (75%), (b) confirmation density algorithm (CDA) setpoints, (c) algorithm tuning parameters, and (d) period based detection algorithm (PBDA),
growth rate algorithm (GRA), and anplitude based algorithm (ABA) setpoints.
The first set, the OPRM auto-enable region setpoints, are treated as nominal setpoints with no additional margins added. The second set, the OPRM CDA trip setpoints, are established in accordance with methodologies defined in Reference 20, and are documented in plant procedures. There are no allowable values for these setpoints. The third set, the OPRM algorithm tuning parameters, are established or adjusted in accordance with and controlled by plant procedures. The fourth set, the PBDA, GRA, and ABA setpoints, are established in accordance with Reference 20 as nominal values only and are controlled by plant procedures.
RPS Instrumentation RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 (continued)
REQU IREMENTS The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing - applicable to Function 2.b and the auto-enable portion of Function 2.f only), the 2-Out-Of-4 Voter channels, and the interface connections into the RPS trip systems from the voter channels.
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The 184-day Frequency of SR 3.3.1.1.11 is based on the reliability analyses of References 15 and 16.
(NOTE:
The actual voting logic of the 2-Out-
-L Function is tested as part of SR 3.3.1.1.15.
e 4:I ;1 o+p.+
IL *
- o.
trp ry confirmed by SR 33111.
A Note is provided for Function 2.a that requires this SR to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entering MODE 2 from MODE 1.
Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads.
This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.
A second Note is provided for Functions 2.b and 2.f that clarifies that the CHANNEL FUNCTIONAL TEST for Functions 2.b and 2.f includes testing of the recirculation flow processing electronics, excluding the flow transmitters.
SR 3.3.1.1.13 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.
This test verifies that the channel responds to the measured parameter within the necessary range and accuracy.
CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
The CHANNEL CALIBRATION for Functions 5 and 8 should consist.of a physical inspection and actuation of the associated position switches.
Note 1 states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.
Changes in neutron detector sensitivity (continued)
Brunswick Unit 1 Revision No.-2,9
.B 3.3-35
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 (continued)
REQUIREMENTS The 24 month Frequency is consistent with the BNP refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.
SR 3.3.1.1.18 The APRM Simulated Thermal Power-High Function (Function 2.b) uses drive flow to vary the trip setpoint.
The OPRM Upscale Function (Function 2.f) uses drive flow to automatically enable or bypass the OPRM Upscale trip output to RPS.
Both of these Functions use drive flow as a representation of reactor core flow.
SR 3.3.1.1.13 assures that the drive flow transmitters and processing electronics are calibrated.
This SR adjusts the recirculation drive flow scaling factors in each APRM channel to provide the appropriate drive flow/core flow alignment.
The Frequency of once within 7 days after reaching equilibrium conditions following a refueling outage is based on the expectation that any change in the core flow to drive flow functional relationship during power operation would be gradual and the maintenance on the Recirculation System and core components which may impact the relationship is expected to be performed during refueling outages.
The 7 day time period after reaching equilibrium conditions is based on plant conditions required to perform the test, engineering judgment of the time required to collect and analyze the necessary flow data, and engineering judgment of the time required to enter and check the applicable scaling factors in each of the APRM channels.
The 7-day time period after reaching equilibrium conditions is acceptable based on the relatively small alignment errors expected, and the margins already included in the APRM Simulated Thermal Power-High and OPRM Upscale Function trip-enable setpoints.
SR
_3.3.4149 euo nbl tpit.Tcaterable setpe~tS.-t1-UUWu-U ab Referencc 21.
This suryeillhrc en~sures that the OPRni (continued)'
Revision No.44_i Brunswick Unit I B 3.3-40
RPS Instrumentation B 3.3.1.1 BASES REQURRIEMNTS I
Ilnrlfr-Tt I
QtjGr nr7lrukr SR 3-3 1...'"t d)
.*ri"- f!...
th e. surve'llan............
a the* AP^
- 4m,,
- ÷,,
-rzPewelr)^
- an r e^--"
a ie e f.--
I-l ow -
properly cVo*r-rclat.e w:ith TrIERPL POWER (SR 3.3.1.1.3) and core flow (SR 3....8,rczpcctivcly.
T
- ya yautoJ enble i
ctpeint is n lnlllslryat iv (i.i, the OPRII Upscale trip is bypassed when APRtI Simulated Tmermal Power it 25% and rccirculaticn driye flow --g ---'),
thrP thee affected channel is..onsidered 4neperable for the OPRM Upscale Function. Alter-natively, the OPRM Upscale trip auto enable setpeintc(s) may be adjusted to plaee the eharmncl in ati'e v.nditicn (not bypassed).
if the
- PRh, Ups.al
.trip is placed in the not bypazzed conditio, thi-S SR is met and the eharnnel is sideed 6PERABE.
The Frcgucrncy ef 24 mcrnth3s 4based on. crgirneerirg judgment.
ind reliability of the eompenent-s.
REFERENCES
- 1.
UFSAR, Section 7.2.
- 2.
UFSAR, Chapter 15.0.
- 3.
UFSAR, Section 7.2.2.
- 4.
NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
- 5.
- 6.
NEDO-23842, Continuous Control Rod Withdrawal in the Startup Range, April 18, 1978.
- 7.
UFSAR, Section 5.2.2.
- 8.
UFSAR, Appendix 5.2A.
- 9.
UFSAR, Section 6.3.1.
(continued)
Brunswick Unit I II B 3.3-41 Revision No.-4r,
RPS Instrumentation B 3.3.1.1 BASES REFERENCES
- 10.
P. Check (NRC) letter to G. Lainas (NRC),
Discharge System Safety Evaluation, December 1, 1980.
- 11.
NEDC-30851-P-A, Technical Specification Improvement Analyses for BWR Reactor Protection System, March 1988.
- 12.
MDE-81-0485, Technical Specification Improvement Analysis for the Reactor Protection System for Brunswick Steam Electric Plant, Units 1 and 2, g
- April 1985.
lc
- 13.
UFSAR, Table 7.2.1-3.
S14.
NEDO-32291-A, System Analyses for the Elimination of o
Selected Response Time Testing Requirements, "1)
October 1995.
10
- 15.
NEDC-32410P-A, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Trip Function, October 1995.
- 16.
NEDC-3241OP-A, Supplement 1, Nuclear Measurement 0
Analysis and Control Power Range Neutron Monitor 1
-j (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function, November 1997.
- 17.
NEDO-31960-A, BWR Owners' Group Long-Term Stability Solutions Licensing Methodology, November 1995.
- 18.
NEDO-31960-A, Supplement 1, BWR Owners' Group Si Long-Term Stability Solutions Licensing Methodology, r
.November 1995.
- Q.2
- 19.
NEDO-32465-A, BWR Owners' Group Long-Term Stability 4
- Detect and Suppress Solutions Licensing Basis 2*Z*cd
.ethodology and Reload Applications, August 1996.
2-0.
Lctter, 6. A. Englan~d (B'WROG) to Wj. Virgilie, BW~R Ownrscr' Group Cuidceline;z for Stability interim 10 Gtrreetiye Aetien, June 6, 1994.
phillip2 (NflC), Cuidolir16-fcr Stability OptiontneI "E
1 nable Region1" (TAG ?192882), September 17, 199.
(continued)
Revision No. 21 I Brunswick Unit I B 3.3-42
RPS Instrumentation B 3.3.1.1 BASES REFERENCES 2&"G2...,,,1 El.t, ez Nul'---r Energy Letter NSA 01 2"
1 (continued) 9RF C51 00251 00, A. rhung (CE) to S.
^ c raort.
(CE), "Minimum Pluffbcr ef Gpcraablc OPRIIl Gells floi Option~ H! Stability at B-run3wiek 1 a~,
dune 0, 2001.
- 23.
NEDE 24011 P At, General rElectric4 Standard JApplication for Road Fuel, (latest approved verlsion).
Brunswick Unit 1 Revision No. q'll B 3.3-43
Recirculation Loops Operating I
B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued)
LCO A plant specific LOCA analysis has been performed assuming only one operating recirculation loop.
This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, without the requirement to modify the APLHGR requirements (Ref. 3).
However, the COLR may require APLHGR limits to restrict the peak clad temperature for a LOCA with a single recirculation loop operating below the corresponding temperature for both loops operating.
The transient analyses of Chapter 15 of the UFSAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed without the requirement to modify the MCPR requirements.
During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM)
Simulated Thermal Power-High Allowable Value is required to account for the different analyzed limits between two-recirculation drive flow loop operation and operation with only one loop.
The APRM channel subtracts the LW value from the measured recirculation drive flow to effectively shift the limits and uses the adjusted recirculation drive flow value to determine the APRM Simulated Thermal Power-High Function trip setpoint.
Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref.
4).
Two recirculation loops are normally required to be in operation with their recirculation pump speeds matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied.
Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"),
MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"),
and APRM Simulated Thermal Power-High (continued)
Brunswick Unit I yq.-
Revi si on No. -pr B 3.4-3
SRVs B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.3 Safety/Relief Valves (SRVs)
BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves.
As part of the nuclear pressure relief system, the size and number of SRVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).
The SRVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell.
The SRVs can actuate by either of two modes:
the safety mode or the relief mode (However, for the purposes of this LCO, only the safety mode is required).
In the safety mode (or spring mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed.
Opening the pilot valve allows a pressure diffdrential to develop across the main valve piston and opens the main valve.
This satisfies the Code requirement.
Each SRV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.
The SRVs that provide the relief mode are the Automatic Depressurization System (ADS) valves.
The ADS requirements are specified in LCO 3.5.1, "ECCS-Operating."
APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressurization transient.
Evaluations have e* ermined atthe most severe
.i is the closure of
/lLsis*
all main steam isolation valves (MSIVs
, followed by reactor s scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). For the purpose of the analyses, 9 SRVs are assumed to operate in the safety mode.
The analysis results demonstrate that the design SRV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel dnesin pressure (110% x 1250 psig = 35pi l*
.~*t
/lllhat 4h.
neeil i*
5*
ii~
I il of5 V
l l
137 SIV 14li1l-l I1 (continued)
Brunswick Unit I B 3.4-11 Revision No.-4&_
SRVs B 3.4.3 BASES APPLICABLE r*-
4verpressu.zatiin-a33c;ted with An ATZP
- Pvtnt, SAFETY ANALYSES 1e assumed to operate in 4n the s*fety m.
- d. 7Th-c^ \\
(continued) analysis.....
- 2),
t
.n.t.,cte ththe
- e**.
cp^ cTTU H!
S nde Scr'...i.
Lc..l C limits..
From an overpressure standpoint, the design basis events are bounded by the overpressurization associated with the ATWS event described above.
Reference 3 discusses additional Sevents that are expected to actuate the SRVs.
SRVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).
LCO The safety function of(USRVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1,
- j3 YOL*'
)
2,'
).
The requirements of this LCO are applicable only to Fh-icapability of the SRVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function).
The SRV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied.
The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions.
These setpoints also ensure that in the event of an ATWS, the reactor pressure remains below the ATWS limit of 1500 psig.
The transient evaluations in the UFSAR involving the safety mode are based on these setpoints, but also include the additional uncertainties of +/- 3% of the nominal setpoint drift to provide an added degree of conservatism.
Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.
APPLICABILITY In MODES 1, 2, and 3, all required SRVs must be OPERABLE, since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in (continued)
Brunswick Unit I Revision No.-ef-*
B 3.4-12
Insert 5 For overpressuization associated with an ATWS event initiated from within the MIELLLA operating region, 10 SRVs operated in the safety mode will maintain reactor pressure below the ASME Section III Code Service Level C limits of 1500 psig (Ref. 2). For overpressuization associated with an ATWS event initiated from within the MELLLA+ operating region, 11 SRVs operated in the safety mode will maintain reactor pressure below the ASME Section III Code Service Level C limits JRef. 6). However, to bound potential surveillance test results, this MELLLA+ case assumed a 10% setpoint drift on a limiting SRV. The results showed that the vessel overpressure criterion was met with this configuration.
Insert 6 This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event and the acceptance limit of 1500 psig is met during an ATWS Event.
SRVs B 3.4.3 BASES APPLICABILITY (continued) these MODES.
The SRVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR)
System is capable of dissipating the core heat.
In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents.
In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure.
The SRV function is not needed during these conditions.
ACTIONS
.th lss then the miniumrnumber *f degu~eJ SR's OPERAB*,
- ransicnt mnay rcsul-t-in the vio-.lation of-the ASME Czde19 lI Id t V1 In sUdu.~
JJ 1F the safety funetion of one or Aoorquie I
R~
I --
oprable, the 'plant miust be brought to a MODE in which the LCO dccz not apply. Te aehic'.t this status,- the plant must be bi-euht to MODE 3F withn 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4I within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The 0alkwec Czmpletien Times are reasonable, bascd-an-opera-4fti e~eeneey to--reach reqv4*red-plant conditions from full power conditions in an orderly manner nfitf~
challenging plýant
.,,.tc.....
SURVEILLANCE SR 3.4.3.1 6i1 REQUIREMENTS This Surveillance requires that the required(
SRVs will o en at the ressures assumed in the safety analysis of 3
eferences 1,,
The demonstration of the SRV safety function lift set ings must be performed during shutdown, since this is a bench test, in accordance with the Inservice Testing Program.
The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
SR 3.4.3.2 A manual actuation of each required SRV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line.
This (continued)
Revision No.*
Brunswick Unit 1 B 3.4-13
Insert 7 In order to ensure that the reactor pressure remains below the ATWS transient limit of 1500 psig, for events initiated while in the MELLLA+ operating region, 11 SRVs must be OPERABLE. If the safety function of one SRV is inoperable, the plant must be brought to a condition in which the LCO does not apply. To achieve this status, the plant must exit the MELLLA+ operating region, as defined in the COLR, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time is reasonable to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
Insert 8 B.1 and B.2 If the safety function of two or more SRVs is inoperable, a transient may result in the violation of reactor vessel pressure ASME Code limits regardless of the operating region from which the event initiated. As such, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SRVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 (continued)
REQU IREM ENTS can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow.
Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve.
Sufficient time is therefore allowed after the required pressure is achieved to perform this test.
Adequate pressure at which this test is to be performed, to avoid damaging the valve, is 945 psig.
Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation.
Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.
The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR.
If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the SRV is considered OPERABLE.
The 24 month Frequency was developed based on the SRV tests required by the ASME Boiler and Pressure Vessel Code,Section XI (Ref.
5).
Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency.
Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES
- 1.
UFSAR, Section 5.2.2.2.
- 2.
NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, Supplement 1, March 1996.
- 3.
UFSAR, Chapter 15.
- 4.
- 5.
ASME, Boiler and Pressure Vessel Code,Section XI.
6E, wD
".3 0
/..'-T1sAA+/--t )cLVyd Brs ckctive Lnrit Iq" B.-4 Revision No.
1 Brunswick Unit 1 B 3.4-14 Revision No.
BSEP 02-0169 Page 1 of 1 List Of Regulatory Commitments The following table identifies those actions committed to by Carolina Power & Light (CP&L)
Company in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to the Manager - Regulatory Affairs at the Brunswick Steam Electric Plant.
n Sh d dl
- 1. None N/A N