BSEP 04-0046, Submittal of Technical Specification Bases Changes
| ML040970397 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/29/2004 |
| From: | O'Neil E Progress Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BSEP 04-0046 | |
| Download: ML040970397 (23) | |
Text
Co Progress Energy MAR 2 9 2004 SERIAL: BSEP 04-0046 U. S. Nuclear Regulatory Commission ATTN: 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 Submittal of Technical Specification Bases Changes Ladies and Gentlemen:
In accordance with Technical Specification (TS) 5.5.10 for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2, Carolina Power & Light Company, now doing business as Progress Energy Carolinas, Inc., is submitting Revisions 33, 34, and 35 to the BSEP, Unit 1 TS Bases and Revision 32 to the BSEP, Unit 2 TS Bases.
Please refer any questions regarding this submittal to Mr. Leonard R. Beller, Supervisor -
Licensing/Regulatory Programs, at (910) 457-2073.
Sincerely, Edward T. O'Neil Manager - Support Services Brunswick Steam Electric Plant MAT/mat
Enclosures:
- 1. Summary of Revisions to Technical Specification Bases
- 2. Page Replacement Instructions
- 3. Unit 1 Technical Specification Bases Replacement Pages
- 4. Unit 2 Technical Specification Bases Replacement Pages Progress Energy Carolinas, Inc.
C( D Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28481
Document Control Desk BSEP 04-0046 / Page 2 cc (with enclosures):
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 ATLN: Mr. Eugene M. DiPaolo, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)
ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Ms. Beverly 0. Hall, Section Chief Radiation Protection Section, Division of Environmental Health North Carolina Department of Environment and Natural Resources 3825 Barrett Drive Raleigh, NC 27609-7221
BSEP 04-0046 Page 1 of 2 Summary of Revisions to Technical Specification Bases Revision Date Implemented Title/Description Revison
-Unit 33 1
January 21, 2004
Title:
Incorporation of COLR Reference 32 2
versus MCPR Value in B 3.3.2.1
==
Description:==
Revision 33 for Unit 1 and 32 for Unit 2 revise B 3.3.2.1, "Control Rod Block Instrumentation," by replacing a cycle specific value for Minimum Critical Power Ratio (MCPR) within the Rod Block Monitor discussion with a reference to the Core Operating Limits Report (COLR).
34 1
March 3,2004
Title:
Unit 1, SLC Sodium Pentaborate Requirements
==
Description:==
Unit 1, Revision 34 incorporates changes to B 3.1.7 "SLC System," associated with License Amendment 227, "Sodium Pentaborate Solution Requirements," issued March 25, 2003, to be incorporated prior to the end of the Unit 1, Cycle 15 refueling outage.
Title:
Clarification to Bases 3.3.1.1, Action I
==
Description:==
Unit 1, Revision 34 also incorporates a clarification to the entry conditions for Action I of Specification 3.3.1.1, "RPS Instrumentation."
BSEP 04-0046 Page 2 of 2 Summary of Revisions to Technical Specification Bases Revision Date Implemented Title/Description nit 35 1
March 27, 2004
Title:
Unit 1, Safety Limit Minimum Critical Power Ratio (SLMCPR)
==
Description:==
Unit 1, Revision 35 incorporates changes to B 3.2.2, "Minimum Critical Power Ratio," associated with License Amendment 231, issued March 26, 2004.
BSEP 04-0046 Page 1 of 1 Page Replacement Instructions
-:Remove -
Insert Unit 1 - 'Bases Book 1:
Cover Page, Revision 32 Cover Page, Revision 35 LOEP-1, Revision 32 LOEP-1, Revision 35 LOEP-2, Revision 31 LOEP-2, Revision 35 B 3.1.7-1, Revision 31 B 3.1.7-1, Revision 34 B 3.1.7-6, Revision 31 B 3.1.7-6, Revision 34 B 3.2.2-1, Revision 31 B 3.2.2-1, Revision 35 B 3.2.2-2, Revision 31 B 3.2.2-2, Revision 35 B 3.2.2-4, Revision 31 B 3.2.2-4, Revision 35 B 3.2.2-5, Revision 31 B 3.2.2-5, Revision 35 B 3.3.1.1-28, Revision 31 B 3.3.1.1-28, Revision 34 B 3.3.1.1-29, Revision 31 B 3.3.1.1-29, Revision 34 B 3.3.2.1-4, Revision 31 B 3.3.2.1-4, Revision 33 Unit 2 - Bases Book i Cover Page, Revision 31 Cover Page, Revision 32 LOEP-1, Revision 31 LOEP-1, Revision 32 LOEP-2, Revision 30 LOEP-2, Revision 32 B 3.3.2.1-4, Revision 30 B 3.3.2.1-4, Revision 32
BSEP 04-0046 Unit 1 Technical Specification Bases Replacement Pages
BASES TO THE FACILITY OPERATING LICENSE DPR-71 TECHNICAL SPECIFICATIONS FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 1 CAROLINA POWER & LIGHT COMPANY REVISION 35
LIST OF EFFECTIVE PAGES - BASES Page No.
Revision No.
Title Page 35 List of Effective Pages - Book 1 LOEP-1 35 LOEP-2 35 LOEP-3 32 LOEP-4 31 i
31 ii 31 B 2.1.1-1 31 B 2.1.1-2 31 B 2.1.1-3 31 B 2.1.1-4 31 B 2.1.1-5 31 B 2.1.2-1 31 B 2.1.2-2 31 B 2.1.2-3 31 B 3.0-1 31 B 3.0-2 31 B 3.0-3 31 B 3.0-4 31 B 3.0-5 31 B 3.0-6 31 B 3.0-7 31 B 3.0-8 31 B 3.0-9 31 B 3.0-10 31 B 3.0-11 31 B 3.0-12 31 B 3.0-13 31 B 3.0-14 31 B 3.0-15 31 B 3.0-16 31 B 3.1.1-1 31 B 3.1.1-2 31 B 3.1.1-3 31 B 3.1.1-4 31 B 3.1.1-5 31 B 3.1.1-6 31 Page No.
B 3.1.2-1 B 3.1.2-2 B 3.1.2-3 B 3.1.2-4 B 3.1.2-5 B 3.1.3-1 B 3.1.3-2 B 3.1.3-3 B 3.1.3-4 B 3.1.3-5 B 3.1.3-6 B 3.1.3-7 B 3.1.3-8 B 3.1.3-9 B 3.1.4-1 B 3.1.4-2 B 3.1.4-3 B 3.1.4-4 B 3.1.4-5 B 3.1.4-6 B 3.1.4-7 B 3.1.5-1 B 3.1.5-2 B 3.1.5-3 B 3.1.5-4 B 3.1.5-5 B 3.1.6-1 B 3.1.6-2 B 3.1.6-3 B 3.1.6-4 B 3.1.6-5 B 3.1.7-1 B 3.1.7-2 B 3.1.7-3 B 3.1.7-4 B 3.1.7-5 B 3.1.7-6 B 3.1.8-1 B 3.1.8-2 B 3.1.8-3 B 3.1.8-4 B 3.1.8-5 Revision No.
31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 34 31 31 31 31 34 31 31 31 31 31 I
I (continued)
Brunswick Unit 1 LOEP-1 Revision 35 l
LIST OF EFFECTIVE PAGES - BASES (continued)
Page No.
Revision No.
Page No.
Revision No.
B 3.2.1-1 B 3.2.1-2 B 3.2.1-3 B 3.2.1-4 B 3.2.1-5 B 3.2.2-1 B 3.2.2-2 B 3.2.2-3 B 3.2.2-4 B 3.2.2-5 B 3.3.1.1 -1 B 3.3.1.1 -2 B 3.3.1.1-3 B 3.3.1.1 -4 B 3.3.1.1-5 B 3.3.1.1-6 B 3.3.1.1 -7 B 3.3.1.1 -8 B 3.3.1.1-9 B 3.3.1.1 -1 0 B 3.3.1.1 -11 B 3.3.1.1-12 B 3.3.1.1-13 B 3.3.1.1-14 B 3.3.1.1-15 B 3.3.1.1-16 B 3.3.1.1-17 B 3.3.1.1-18 B 3.3.1.1-19 B 3.3.1.1-20 B 3.3.1.1-21 B 3.3.1.1-22 B 3.3.1.1-23 B 3.3.1.1-24 B 3.3.1.1-25 B 3.3.1.1-26 B 3.3.1.1-27 B 3.3.1.1-28 B 3.3.1.1-29 B 3.3.1.1-30 B 3.3.1.1-31 B 3.3.1.1-32 B 3.3.1.1-33 31 31 31 31 31 35 35 31 35 35 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 34 34 31 31 31 31 B 3.3.1.1-34 B 3.3.1.1-35 B 3.3.1.1-36 B 3.3.1.1-37 B 3.3.1.1-38 B 3.3.1.1-39 B 3.3.1.1-40 B 3.3.1.1-41 B 3.3.1.1-42 B 3.3.1.1-43 B 3.3.1.2-1 B 3.3.1.2-2 B 3.3.1.2-3 B 3.3.1.2-4 B 3.3.1.2-5 B 3.3.1.2-6 B 3.3.1.2-7 B 3.3.1.2-8 B 3.3.1.2-9 B 3.3.2.1 -1 B 3.3.2.1-2 B 3.3.2.1-3 B 3.3.2.1-4 B 3.3.2.1-5 B 3.3.2.1-6 B 3.3.2.1-7 B 3.3.2.1-8 B 3.3.2.1-9 B 3.3.2.1-10 B 3.3.2.1 -11 B 3.3.2.1-12 B 3.3.2.1-13 B 3.3.2.1-14 B 3.3.2.1-15 B 3.3.2.2-1 B 3.3.2.2-2 B 3.3.2.2-3 B 3.3.2.2-4 B 3.3.2.2-5 B 3.3.2.2-6 B 3.3.2.2-7 B 3.3.3.1-1 B 3.3.3.1-2 B 3.3.3.1-3 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 33 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 I
I (continued)
Brunswick Unit 1 LOEP-2 Revision 35 l
SLC System B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10 CFR 50.62 (Ref. 1) on anticipated transient without scram.
The SLC System is also used to maintain suppression pool pH level above 7 following a loss of coolant accident (LOCA) involving significant fission product releases. Maintaining suppression pool pH levels greater than 7 following an accident ensures that iodine will be retained in the suppression pool water (Ref. 2).
The SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized water is provided for testing purposes.
APPLICABLE The SLC System is manually initiated from the main control room, as SAFETY ANALYSES directed by the emergency operating procedures, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods.
The SLC System is used in the event that enough control rods cannot be inserted to accomplish shutdown and cooldown in the normal manner.
The SLC System injects borated water Into the reactor core to add negative reactivity to compensate for all of the various reactivity effects that could occur during plant operations. To meet this objective, it is necessary for SLC to inject a quantity of boron which produces a concentration of 720 ppm equivalent of natural boron in the reactor coolant at 700F with normal reactor vessel water level. To allow for (continued)
Brunswick Unit 1 B 3.1.7-1 Revision No. 34 l
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 (continued)
REQUI REMENTS that has been certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 48 months at alternating 24 month intervals. The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated 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.
SR 3.1.7.8 Enriched sodium pentaborate solution is made by mixing granular, enriched sodium pentaborate with water. Isotopic tests on the granular sodium pentaborate to verify the actual B-1 0 enrichment must be performed prior to addition to the SLC tank in order to ensure that the proper B-10 atom percentage is being used.
REFERENCES
- 1.
- 2.
NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, Final Report, February 1, 1995.
- 3.
UFSAR, Section 9.3.4.
- 4.
- 5.
Brunswick Unit 1 B 3.1.7-6 Revision No. 34 I
MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref.1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.
The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.
APPLICABLE The analytical methods and assumptions used in evaluating the AOOs SAFETY ANALYSES to establish the operating limit MCPR are presented in References 2, 3, 4, 5, 6, 7, and 8. To ensure that the 99.9% of the fuel rods avoid boiling transition during any transient that occurs with moderate frequency, limiting transients are analyzed either with TRACG or other methodologies. The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The TRACG methodology calculates an operating limit MCPR (OLMCPR) for the transient initial condition that will yield the largest change in CPR (ACPR) resulting from the limiting transient. When the largest ACPR is added to the MCPR SL, an OLMCPR is obtained.
The most limiting of the OLMCPR calculated by either the TRACG or other methodology sets the core operating limits.
The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPRj and (continued)
Brunswick Unit 1 B 3.2.2-1 Revision No. 35 l
MCPR B 3.2.2 BASES APPLICABLE MCPRp, respectively) to ensure adherence to fuel design limits during the SAFETY ANALYSES worst transient that occurs with moderate frequency (Ref. 7).
(continued)
Flow dependent MCPR limits are determined using the methodology described in Reference 2 to analyze slow flow runout transients. The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.
Power dependent MCPR limits (MCPRp) are determined using the methodology described in Reference 2. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPRp operating limits are provided for operating between 23% RTP and the previously mentioned bypass power level.
The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 9).
I LCO The MCPR operating limits, as a function of core flow, core power, and cycle exposure, specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the MCPRf and MCPRp limits.
APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 23% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 23% RTP is unnecessary due to the large inherent margin that ensures that the MCPRSL is not exceeded even if a limiting transient occurs.
Statistical analyses indicate that the nominal value of the initial MCPR expected at 23% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as (continued)
.Brunswick Unit 1 B 3.2.2-2 Revision No. 35 l
MCPR B 3.2.2 BASES SURVEILLANCE SR 3.2.2.1 (continued)
REQUIREMENTS slowness of changes in power distribution during normal operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER 2 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.
SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis.
SR 3.2.2.2 determines the value of T, which is a measure of the actual scram speed distribution compared with the assumed distribution. The MCPR operating limit is then determined based on an interpolation between the applicable limits for Option A (scram times of LCO 3.1.4,
'Control Rod Scram Times") and Option B (realistic scram times) analyses. The MCPR operating limits for the Option A and Option B analyses are specified in the COLR. The parameter T must be determined once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is acceptable due to the relatively minor changes in T expected during the fuel cycle.
REFERENCES
- 1.
UFSAR Section 4.4.2.1.
- 2.
NEDO-2401 1-P-A, General Electric Standard Application for Reactor Fuel (latest approved version).
- 3.
UFSAR, Chapter 4.
- 4.
UFSAR, Chapter 6.
- 5.
UFSAR, Chapter 15.
- 6.
NEDC-31776P, Brunswick Steam Electric Plant Units 1 and 2 Single-Loop Operation, December 1989.
(continued)
Brunswick Unit I B 3.2.2-4 Revision No. 35 l
MCPR B 3.2.2 BASES REFERENCES
- 7.
NEDC-31654P, Maximum Extended Operating Domain (continued)
Analysis for Brunswick Steam Electric Plant, February 1989.
- 8.
NEDE-32906P-A, "TRACG Application for Anticipated Operational Occurrences (AOO) Transient Analyses,' approved version as specified in the COLR.
- 9.
Brunswick Unit 1 B 3.2.2-5 Revision No. 35 l
RPS Instrumentation B 3.3.1.1 BASES ACTIONS D.1 (continued)
Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition.
E.1, F.1, G.1, and J.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Actions E.1 and J.1 are consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR).'
H.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
1.1 Condition I exists when the OPRM Upscale Trip capability has been lost for all APRM channels due to unanticipated equipment design or Instability detection algorithm problems, or when testing to confirm plant response/performance following algorithm modifications.
References 15 and 16 justified use of alternate methods to detect and suppress oscillations under limited conditions. The alternate methods are procedurally established consistent with the guidelines identified in Reference 20. The alternate (continued)
Brunswick Unit 1 B 3.3.1.1-28 Revision No. 34
RPS Instrumentation B 3.3.1.1 BASES ACTIONS 1.1 (continued) methods procedures require operating outside a "restricted zone" in the power-flow map and manual operator action to scram the plant if certain predefined events occur. The 12-hour allowed Completion Time for Required Action 1.1 is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place. Based on the small probability of an instability event occurring at all, the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is judged to be reasonable.
This Required Action is intended to allow continued plant operation under limited conditions when an unanticipated equipment design or instability detection algorithm problem causes OPRM Upscale Function inoperability in all APRM channels. This Required Action is not intended and was not evaluated as a routine alternative to return failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to be accomplished within the completion times allowed for Required Actions for Condition A. The alternate method to detect and suppress oscillations implemented in accordance with 1.1 is intended to be applied only as long as is necessary to implement and test corrective action to resolve the unanticipated equipment design or instability detection algorithm problem.
SURVEILLANCE As noted at the beginning of the SRs, the SRs for each RPS REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.1.1-1.
The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, provided the associated Function maintains RPS trip capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 11, 15, and 16) assumption of the average time required to perform channel Surveillance.
(continued)
Brunswick Unit 1 B 3.3.1.1-29 Revision No. 34 l
Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE
- 1. Rod Block Monitor (continued)
SAFETY ANALYSES, LCO, and uncertainties, process effects, calibration tolerances, instrument drift, and APPLICABILITY environment errors are accounted for and appropriately applied for the instrumentation.
The RBM is assumed to mitigate the consequences of an RWE event when operating 2 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE (Ref. 2). When operating < 90% RTP, analyses (Ref. 2) have shown that with an initial MCPR 2 1.70, no RWE event will result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at > 90% RTP with MCPR greater than or equal to the limit specified in the COLR, no RWE event will result in exceeding the MCPR SL (Ref. 2). Therefore, under these conditions, the RBM is also not required to be OPERABLE.
The RBM selects one of three different RBM flux trip setpoints to be applied based on the current value of THERMAL POWER. THERMAL POWER is indicated to each RBM channel by a simulated thermal power (STP) reference signal input from an associated reference APRM channel. The OPERABLE range is divided into three "power ranges," a "low power range," an "intermediate power range," and a "high power range." The RBM flux trip setpoint applied within each of these three power ranges is, respectively, the "low trip setpoint," the "intermediate trip setpoint," and the "high trip setpoint" (Allowable Values for which are defined in the COLR). To determine the current power range, each RBM channel compares its current STP input value to three power setpoints, the "low power setpoint" (29%), the "intermediate power setpoint" (current value defined in the COLR), and the "high power setpoint" (current value defined in the COLR), which define, respectively, the lower limit of the low power range, the lower limit of the intermediate power range, and the lower limit of the high power range. The trip setpoint applicable for each power range is more restrictive than the corresponding setpoint for the lower power range(s). When STP is below the low power setpoint, the RBM flux trip outputs are automatically bypassed but the low trip setpoint continues to be applied to indicate the RBM flux setpoint on the NUMAC RBM displays.
(continued)
Brunswick Unit I B 3.3.2.1-4 Revision No. 33 l
BSEP 04-0046 Unit 2 Technical Specification Bases Replacement Pages
BASES TO THE FACILITY OPERATING LICENSE DPR-62 TECHNICAL SPECIFICATIONS FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 CAROLINA POWER & LIGHT COMPANY REVISION 32
LIST OF EFFECTIVE PAGES - BASES Page No.
Revision No.
Title Page 32 List of Effective Pages - Book 1 LOEP-1 32 LOEP-2 32 LOEP-3 31 LOEP-4 30 i
30 ii 30 B 2.1.1 -1 30 B 2.1.1-2 30 B 2.1.1-3 30 B 2.1.1-4 30 B 2.1.1-5 30 B 2.1.2-1 30 B 2.1.2-2 30 B 2.1.2-3 30 B 3.0-1 30 B 3.0-2 30 B 3.0-3 30 B 3.0-4 30 B 3.0-5 30 B 3.0-6 30 B 3.0-7 30 B 3.0-8 30 B 3.0-9 30 B 3.0-10 30 B 3.0-11 30 B 3.0-12 30 B 3.0-13 30 B 3.0-14 30 B 3.0-15 30 B 3.0-16 30 B 3.1.1-1 30 B 3.1.1-2 30 B 3.1.1-3 30 B 3.1.1-4 30 B 3.1.1-5 30 B 3.1.1-6 30 Page No.
B 3.1.2-1 B 3.1.2-2 B 3.1.2-3 B 3.1.2-4 B 3.1.2-5 B 3.1.3-1 B 3.1.3-2 B 3.1.3-3 B 3.1.3-4 B 3.1.3-5 B 3.1.3-6 B 3.1.3-7 B 3.1.3-8 B 3.1.3-9 B 3.1.4-1 B 3.1.4-2 B 3.1.4-3 B 3.1.4-4 B 3.1.4-5 B 3.1.4-6 B 3.1.4-7 B 3.1.5-1 B 3.1.5-2 B 3.1.5-3 B 3.1.5-4 B 3.1.5-5 B 3.1.6-1 B 3.1.6-2 B 3.1.6-3 B 3.1.6-4 B 3.1.6-5 B 3.1.7-1 B 3.1.7-2 B 3.1.7-3 B 3.1.7-4 B 3.1.7-5 B 3.1.7-6 B 3.1.8-1 B 3.1.8-2 B 3.1.8-3 B 3.1.8-4 B 3.1.8-5 Revision No.
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I (continued)
Brunswick Unit 2 LOEP-1 Revision 32 l
LIST OF EFFECTIVE PAGES - BASES (continued)
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Revision No.
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B 3.2.1-1 B 3.2.1-2 B 3.2.1-3 B 3.2.1-4 B 3.2.1-5 B 3.2.2-1 B 3.2.2-2 B 3.2.2-3 B 3.2.2-4 B 3.2.2-5 B 3.3.1.1-1 B 3.3.1.1-2 B 3.3.1.1-3 B 3.3.1.1-4 B 3.3.1.1-5 B 3.3.1.1-6 B 3.3.1.1-7 B 3.3.1.1-8 B 3.3.1.1-9 B 3.3.1.1-10 B 3.3.1.1-11 B 3.3.1.1-12 B 3.3.1.1-13 B 3.3.1.1-14 B 3.3.1.1-15 B 3.3.1.1-16 B 3.3.1.1-17 B 3.3.1.1-18 B 3.3.1.1-19 B 3.3.1.1-20 B 3.3.1.1-21 B 3.3.1.1-22 B 3.3.1.1-23 B 3.3.1.1-24 B 3.3.1.1-25 B 3.3.1.1-26 B 3.3.1.1-27 B 3.3.1.1-28 B 3.3.1.1-29 B 3.3.1.1-30 B 3.3.1.1-31 B 3.3.1.1-32 B 3.3.1.1-33 B 3.3.1.1-34 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 B 3.3.1.1-35 B 3.3.1.1-36 B 3.3.1.1-37 B 3.3.1.1-38 B 3.3.1.1-39 B 3.3.1.140 B 3.3.1.1-41 B 3.3.1.1-42 B 3.3.1.1-43 B 3.3.1.2-1 B 3.3.1.2-2 B 3.3.1.2-3 B 3.3.1.2-4 B 3.3.1.2-5 B 3.3.1.2-6 B 3.3.1.2-7 B 3.3.1.2-8 B 3.3.1.2-9 B 3.3.1.3-1 B 3.3.1.3-2 B 3.3.1.3-3 B 3.3.1.3-4 B 3.3.1.3-5 B 3.3.1.3-6 B 3.3.1.3-7 B 3.3.1.3-8 B 3.3.1.3-9 B 3.3.2.1-1 B 3.3.2.1-2 B 3.3.2.1-3 B 3.3.2.1-4 B 3.3.2.1-5 B 3.3.2.1-6 B 3.3.2.1-7 B 3.3.2.1-8 B 3.3.2.1-9 B 3.3.2.1-10 B 3.3.2.1-11 B 3.3.2.1-12 B 3.3.2.1-13 B 3.3.2.1-14 B 3.3.2.1-15 B 3.3.2.2-1 B 3.3.2.2-2 B 3.3.2.2-3 30 30 30 30 30 30 30 30 30 30 30 30 32 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 I
(continued)
Brunswick Unit 2 LOEP-2 Revision 32 l
Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE
- 1. Rod Block Monitor (continued)
SAFETY ANALYSES, LCO, and uncertainties, process effects, calibration tolerances, instrument drift, and APPLICABILITY environment errors are accounted for and appropriately applied for the instrumentation.
The RBM is assumed to mitigate the consequences of an RWE event when operating 2 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE (Ref. 2). When operating < 90% RTP, analyses (Ref. 2) have shown that with an initial MCPR 2 1.70, no RWE event will result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at 2 90% RTP with MCPR greater than or equal to the limit specified in the COLR, no RWE event will result in exceeding the MCPR SL (Ref. 2). Therefore, under these conditions, the RBM is also not required to be OPERABLE.
The RBM selects one of three different RBM flux trip setpoints to be applied based on the current value of THERMAL POWER. THERMAL POWER is indicated to each RBM channel by a simulated thermal power (STP) reference signal input from an associated reference APRM channel. The OPERABLE range is divided into three "power ranges," a "low power range," an "intermediate power range," and a "high power range." The RBM flux trip setpoint applied within each of these three power ranges is, respectively, the "low trip setpoint," the "intermediate trip setpoint," and the "high trip setpoint" (Allowable Values for which are defined in the COLR). To determine the current power range, each RBM channel compares its current STP input value to three power setpoints, the "low power setpoint" (29%), the "intermediate power setpoint" (current value defined in the COLR), and the "high power setpoint" (current value defined in the COLR), which define, respectively, the lower limit of the low power range, the lower limit of the intermediate power range, and the lower limit of the high power range. The trip setpoint applicable for each power range is more restrictive than the corresponding setpoint for the lower power range(s). When STP is below the low power setpoint, the RBM flux trip outputs are automatically bypassed but the low trip setpoint continues to be applied to indicate the RBM flux setpoint on the NUMAC RBM displays.
J-I Brunswick Unit 2 B 3.3.2.1-4 Revision No. 32 l