Text

Technical Specifications Change TS-443 - Request for Additional Information (RAI) Regarding Oscillation Power Range Monitor
	ML062790140
Person / Time
Site:	Browns Ferry
Issue date:	10/02/2006
From:	Crouch W; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TAC MC9565, TVA-BFN-TS-443
	Download: ML062790140 (71)
	v • d • e

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 October 2, 2006 TVA-BFN-TS-443 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop OWFN, P1-35 Washington, D. C. 20555-0001 Gentlemen:

In the Matter of ) Docket No. 50-259 Tennessee Valley Authority BROWNS FERRY NUCLEAR PLANT (BFN) - UNIT 1 - TECHNICAL SPECIFICATIONS (TS) CHANGE TS-443 - REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING OSCILLATION POWER RANGE MONITOR (OPRM) - (TAC NO. MC9565)

By letter dated January 6, 2006, TVA submitted a request for a Unit 1 TS change (TS-443) to activate the OPRM system (ADAMS Accession No. ML060180477). On August 14, 2006, NRC staff issued a set of RAI questions on TS-443 (ML062220537). to this letter provides TVA's responses to the subject RAI.

TVA has determined that the additional information provided by this letter does not affect the no significant hazards considerations associated with the proposed TS-443 change.

The proposed TS changes still qualify for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).

7!ý) 2~

U.S. Nuclear Regulatory Commission Page 2 October 02, 2006 No new regulatory commitments have been made in this submittal. If you have any questions regarding this letter, please contact William D. Crouch at (256)729-2636.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 2nd day of October, 2006.

Sincerely, William D. Crouch Manager of Licensing and Industry Affairs

Enclosures:

1. Response to Request for Additional Information

2. Copy of BFN Unit 1 TS Section 3.3.1.1 and Bases cc: See page 3

U.S. Nuclear Regulatory Commission Page 3 October 02, 2006

Enclosures:

cc (Enclosures):

State Health Officer Alabama Dept. of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, AL 36130-3017 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970 NRC Unit 1 Restart Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970 Margaret Chernoff, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 1 TS-443 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING OSCILLATION POWER RANGE MONITOR (OPRM)

This enclosure provides TVA's response to NRC's August 14, 2006, RAI (ADAMS Accession No. ML062220537) on BFN Unit 1 Technical Specifications (TS) change No. 443. The RAI questions are repeated below.

NRC RAI 1 Provide the analysis or reference document that links the configuration of the Unit 2 system to the configuration of the system that will be activated for Unit 1. This analysis should include any unit and cycle specific changes and the justifications that these changes do not affect the Unit 1 system capability to meet the system design requirements.

TVA Reply to RAI 1 Unit 1 will implement the same Long Term Stability (LTS) solution as that currently installed and in use on Units 2 and 3. This is commonly referred to as the LTS Option III solution as described in NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications." This solution is provided by the OPRM module of the Power Range Monitoring System (PRNM). The PRNM system is a General Electric (GE) NUMAC system, which is being installed on Unit 1 as a power monitoring instrumentation upgrade and to add the OPRM function for the LTS. A license amendment in support of the Unit 1 PRNM modification was submitted on November 10, 2003 (ML033300129) as TS change TS-430. PRNM/OPRM systems have been in-service on Units 2 and 3 for several years.

The Option III algorithms described in NEDO-32465-A are: the Period Based Detection Algorithm (PBDA), Growth Rate Algorithm E1-I

(GRA), and Amplitude Based Algorithm (ABA). These algorithms are used by the Units 2 and 3 OPRM systems. The Unit 1 OPRM differs from the Units 2 and 3 OPRM systems in that it is based upon Detect and Suppress Solution/Confirmation Density (DSS/CD) software. For Unit 1, however, the base Option III algorithms, PBDA, GRA, and ABA, are retained in DSS/CD as Defense-In-Depth (DID) features. The DSS/CD solution layers another oscillation detection algorithm on top of the existing Option III algorithms. For BFN Unit 1 at Power Uprate/Maximum Extended Load Line Limit Analysis (MELLLA) conditions (105% of original licensed power or 3458 MWt) or at Extended Power Uprate (EPU)/MELLLA conditions (120% of original licensed power or 3952 MWt), the trip from this additional CD algorithm will not be enabled. That is, the CD portion will be operated in "indicate-only" mode. The trips for the remaining three algorithms (PBDA, GRA, and ABA) will be enabled and the Unit 1 OPRM will function as an Option III LTS system. This is similar to the configuration currently in use at the Brunswick plant.

GE performed an assessment of the use of the DID configuration of DSS/CD to meet Option III requirements for BFN Unit 1. In the Reference 1-1 report, GE compared the various OPRM system settings between Option III and DSS/CD-based systems. The primary algorithm of interest is the PBDA since it performs the licensing basis protection of the Safety Limit Minimum Critical Power Ratio (SLMCPR) under Option III. This GE evaluation resulted in the recommendation to change one of the DSS/CD base settings, Tmin, in order to minimize the potential for spurious scrams.

Table 1-1 below compares the primary non-cycle specific licensing basis algorithm (PBDA) settings between the current Unit 2/3 Option III systems to the Unit 1 DSS/CD-based system.

The primary difference is that DSS/CD uses a higher Tmax setting, which effectively increases the sensitivity of the system by increasing the size of the window used for period confirmation. This is a conservative change when compared to the existing Unit 2/3 Option III settings.

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Table 1-1: Comparison of Non-Cycle Specific PBDA Settings Ul U2/U3 DSS/CD Option III Option III (DSS/CD System)

Tmin 1.2 sec 0.8 sec 1.2 sec Tmax 3.0 sec 4.0 sec 4.0 sec Cutoff Freq, Fc 1 Hz 1 Hz 1 Hz Period Terac, 100 msec 100 msec 100 msec Tolerance, With these settings, the Unit 1 OPRM system provides the equivalent function of the Units 2 and 3 OPRM systems in providing the LTS for Unit 1.

Reference:

1-I GE-NE-0000-0047-1802-RO, "Tennessee Valley Authority, Browns Ferry Unit 1, Detect and Suppress Solution -

Confirmation Density, Defense-in-Depth Detection Algorithms Assessment," dated January 2006.

NRC RAI 2 In the Extended Power Uprate (EPU) Technical Specification (TS) change request (TS-431) Tennessee Valley Authority requested an EPU from 3293 to 3952 (approximately 20.012 percent) megawatts thermal (MWt). The General Electric (GE) Licensing Topical Reports (LTRs) NEDO-32465-A and NEDC-32410P-A identify the power threshold to be 30 percent which corresponds to 987.9 MWt

(.3x3293=987.9). The 25 percent rated thermal power threshold requested in TS-443 (ML060180477) is 988 MWt (.25x3952=988) which is slightly nonconservative when compared to that of the approved LTR. Provide justification or explanation.

TVA Reply to RAI 2 The OPRM trip enabled region at EPU/MELLLA conditions (without MELLLA+) is specified in GE Task Report T0202, "Thermal-Hydraulic Stability," which was prepared by GE for TVA (Reference 2-1). For operation at EPU conditions, the Option III trip enabled region was rescaled to maintain the same E1-3

absolute power/flow region boundaries. Because the rated core flow is not changed, the 60% core flow boundary was not rescaled.

The 30% power boundary for OPRM bypass based on Original Licensed Thermal Power (OLTP) boundary was changed by the following equation for EPU:

EPU Region Boundary = 30%OLTP * (100%/EPU(%OLTP))

For 120% EPU:

EPU Region Boundary = 30%OLTP *(100%/120%) = 25% @ EPU This calculation was made based on a rounded power uprate value of 120%. The use of a rounded 120% power uprate value instead of 120.012% does not significantly change the calculated setpoints (25% versus 24.9975%). Although slightly non-conservative, this is considered acceptable.

Reference:

2-1 GE-NE-0000-0015-9413, Revision 0, Task Report T0202, "Thermal-Hydraulic Stability."

NRC RAI 3 The oscillation power range monitor (OPRM) function trips the plant when certain conditions are reached. Provide documentation on the method for determining the allowable value for this trip condition. Describe how the allowable values are documented and implemented.

TVA Reply to RAI 3 The evaluation used in the determination of the Unit 1 Cycle 7 OPRM setpoints for the Option III stability solution is documented in Reference 3-1, Supplemental Reload Licensing Report (SRLR), which was transmitted to NRC on May 15, 2006 (ML061450390). The method to determine the OPRM setpoints (References 3-2 and 3-3) has features which result in significant conservatisms in the determination of the OPRM oscillation amplitude setpoint. These features include:

1) Placing the initial Minimum Critical Power Ratio (MCPR) on the operating limits. Crediting the random errors in the calculated MCPR for the limiting fuel bundle increases the statistically nominal initial MCPR value by several percent.

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2) Calculation of the hot channel response (hot channel oscillation magnitude or HCOM) for anticipated oscillations applies the bounding channel response. An HCOM value based on a cycle-specific statistical upper bound will result in relaxed OPRM setpoints for a given MCPR value.

3) Calculation of the regional DIVOM (which is defined as the Delta CPR over Initial MCPR Versus the Oscillation Magnitude) curve representing a bounding value. A statistically based DIVOM upper bound would result in relaxed OPRM setpoints for a given MCPR value.

The above aspects of the evaluation method result in significant conservatism in the OPRM setpoint even though it does not include allowance for instrument errors. Also note that a nominal setpoint is applied in the methodology for the Option I-D long-term solution.

Instrument errors (bias and random) can cause a difference between the measured amplitude of the OPRM response and the amplitude of the local power oscillation. The instrument errors are due to the specified Local Power Range Monitor (LPRM) detector non-linearity and the specified accuracy of the analog front end of the OPRM electronics that interfaces with the LPRM detectors. Errors due to the rest of the signal processing are negligible because the processing is done digitally with high resolution.

The overall margin required for instrument error is determined from a combination of the LPRM detector and electronic errors.

This margin has been calculated according to GE setpoint methodology and the results show that the bounding instrument error margin in the 30% to 80% core power operating range is less than 1.7% (or 0.017 on the OPRM normalized scale) for OPRM setpoints that are 1.15 or less. Thus, a best estimate instrument error is of the order of 1% for these OPRM setpoints, and this margin can easily be accommodated by the bounding aspects of the evaluation discussed previously.

For example, the conservatisms in two features of Option III are sufficient to offset the OPRM setpoint uncertainty on a generic basis and they cover the three BFN units:

1. The current licensing basis HCOM values represent the 95%

confidence level/95% probability level values. If a nominal HCOM value (average HCOM) was used, the OPRM amplitude setpoint impact would have been reduced by about 0.01.

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Thus, the 95/95 HCOM value provides a 0.01 margin in the OPRM setpoint.

2. The current licensing basis DIVOM slope is calculated based on the limiting power/flow point on the power/flow map. For any two recirculation pump trip (2PT) runback to natural circulation, the power is usually about 10% below the limiting power/flow point. As an example, a post-2PT runback for a plant operating at rated power/MELLLA boundary would end up at a power point roughly halfway between the Extended Load Line Limit Analysis (ELLLA) and MELLLA lines on the natural circulation flow. A more realistic DIVOM slope (shallower) will result in 0.01 margin to the OPRM amplitude setpoint.

Therefore, the conservatisms in the current Option III methodology adequately support the use of nominal setpoints.

References:

3-1 0000-0043-8325-SRLR, Revision 1, Supplemental Reload Licensing Report for Browns Ferry 1 Reload 6 Cycle 7, May 2006.

3-2 NEDO-32465-A, Licensing Topical Report, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

3-3 "Plant-Specific Regional Mode DIVOM Procedure Guideline,"

GE-NE-0000-0028-9714-RI, June 2005.

NRC RAI 4 Describe what TS Bases changes will be made as a result of the activation of the OPRM. Are they consistent with the LTRs?

Justify any deviations.

TVA Reply to RAI 4 Installation of the PRNM system is a prerequisite to implementing the OPRM system. As noted in the response to RAI 1, a license amendment in support of the Unit 1 PRNM modification was submitted to NRC on November 10, 2003 (ML033300129) as TS change TS-430. The TS Bases associated with TS-443 (OPRM) will be integrated into the PRNM TS-430 Bases changes. Hence, after NRC approval of TS-430, the Unit 1 TS Bases will be revised in accordance with TS 5.5.10, TS Bases E1-6

Control Program to incorporate the TS-430 Bases changes. After the subsequent approval of TS-443, the Unit 1 Bases for TS-443 will be integrated into the TS-430 modified TS Bases. Due to the time sequencing of these TS changes, TVA has not yet prepared the specific Unit 1 TS Bases changes that will be made for TS-443.

OPRM TS changes and the associated TS Bases changes for the Unit 2 OPRM installation were submitted on September 8, 1998, in TS change request 354. TS-354 was approved by NRC on March 5, 1999. Unit 3 OPRM TS changes (TS-398) were submitted on July 28, 1999, and the corresponding license amendment was issued by the NRC on September 27, 1999 (ML020100376). The OPRM TS Bases changes for the Units 2 and 3 OPRM TS changes were included in the TS-354 and TS-398 submittals, and were subsequently reissued by NRC to BFN in the referenced safety evaluations. After NRC approval of TS-443, the Unit 1 TS Bases revision will use the TS Bases changes from TS-354 and TS-398 as a blueprint for making the TS-443 Bases changes. The TS Bases from TS-354 and TS-398 closely follow the sample TS Bases in NEDC-32410P-A Supplement 1 (Reference 4-1). Note that TS Bases provide explanatory information for the associated TS provisions and do not alter the requirements of the TS.

Reference:

4-1 Licensing Topical Report NEDC-32410P-A, Supplement 1, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Plus Option III Stability Trip Function.

NRC RAI 5 Since this TS change (TS-443) involves the activation of the OPRM without a trial period as described in the GE LTR, describe related experiences with Units 2 and 3, as well as provide justification for this deviation from the approved plans.

TVA Reply to RAI 5 Following installation of the OPRM systems on Units 2 and 3, the systems were operated for a trial period of one operating cycle with the trip functions disarmed. This approach was recommended by the vendor to allow observation of system response during actual reactor operation. Some false alarms were observed when operating in the OPRM enabled power range.

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Allowable tuning adjustments were made to the Units 2 and 3 OPRM systems to reduce the frequency of the false alarm signals. Because of the similarities between Units 1, 2, and 3, similar tuning adjustments will be made on the Unit 1 OPRM system to minimize false alarms.

Units 2 and 3 have been now been operating with the OPRM Option III system enabled since the Springs of 1999 and 2000, respectively. Thus, BFN has considerable operating experience with the OPRM system in service and OPRM operating characteristics at power are well understood.

The Unit 1 0PRM system differs from Units 2 and 3 in that the Unit 1 OPRM Option III trips are satisfied by the DID algorithms using DSS/CD (see the response to RAI 1). In GE-NE-0000-0047-1802-RO (Reference 1-1), GE evaluated the use of the DID configuration of DSS/CD to meet Option III requirements for BFN Unit 1. This evaluation documented that the primary differences between the Unit 1 and the Units 2 and 3 systems are the wider ranges between the Tmin and Tmax.

These Unit 1 settings increase the sensitivity of the system by increasing the size of the window used for period confirmation.

This is a conservative change when compared to the existing Units 2 and 3 Option III settings and is similar to the system currently installed and operating at Brunswick. Therefore, operating with the Unit 1 OPRM Option III trip enabled is conservative. Considering BFN's extensive operating experience with OPRM in-service on Units 2 and 3, in TS-443, TVA proposed to forego a trial period of operation and that the OPRM be armed at Unit 1 restart, which allows Unit 1 to directly fulfill commitments on implementation of the LTS.

NRC RAI 6 Provide the entire proposed TS section 3.3.1.1 and the associated Bases section. Provide justification for any differences in the surveillance requirements for 2.f of Table 3.3.1.1-1 of the TS and that of NEDC-32410P-A Supplement 1 (see page H-10).

TVA Reply to RAI 6 Installation of the PRNM system is a prerequisite to implementing the OPRM system. As noted in the response to RAI 1, a license amendment in support of the PRNM modification was submitted on November 10, 2003 (ML033300129) as TS change TS-430. Therefore, the TS changes proposed in the January 6, E1-8

2006, Unit 1 OPRM TS-443 submittal were marked-up against Unit 1 TS pages presuming approval of the predecessor PRNM TS change. Likewise, the associated TS-443 OPRM TS Bases changes will be integrated into the TS-430 PRNM TS Bases changes.

Since installation of the PRNM system is a prerequisite to implementing the OPRM system, TVA will first implement TS-430 (after NRC approval) and update the TS Bases for TS-430. Then after approval of TS-443, the OPRM Bases changes will be made in accordance with TS 5.5.10, TS Bases Control Program. As explained in the response to RAI 4, the TS-443 TS Bases changes have not yet been formulated, but will be similar to those provided to NRC for Units 2 and 3 shown in TS-354 and TS-398.

The Bases for TS-398, which were submitted after conversion to standard TS, will be used as the primary blueprint for the Unit 1 TS-443 OPRM TS Bases changes.

A comparison of the differences in the surveillance requirements (SRs) for the changes made in TS-443 to TS 2.f of Table 3.3.1.1-1 of BFN TS and those from NEDC-32410P-A Supplement 1 is shown in the below table:

Table 6-1 Comparison of NEDC-32410P-A SRs Versus TS-443 SRs NEDC-32410P-A SR Description Equivalent Comparison SR Number BFN SR Number 3.1.1.1.1 Channel Check 3.1.1.1.1 12-hour frequency versus BFN 24-hour frequency 3.1.1.1.8 Calibrate LPRMS 3.1.1.1.7 No differences 3.1.1.1.11 Channel 3.1.1.1.16 No differences Functional Test 3.1.1.1.13 Channel 3.1.1.1.13 18-month SR frequency Calibration versus BFN 24-month frequency 3.1.1.1.18 OPRM Not 3.1.1.1.17 NEDC uses 30% as Bypassed bypass point. BFN TS are more conservative.

BFN TS use 25% for scaling as described in RAI 2 response.

Also, 10-month SR frequency versus 24-month frequency.

As can be seen in the table, the TS-443 SRs match those in the NEDC except for differences in performance frequency for the Channel Check and Channel Calibration SRs and for the power E1-9

scaling factor discussed in the RAI 2 response. The SR frequency differences are the same as those issued by NRC for the Units 2 and 3 OPRM TS changes, TS-354 and TS-398.

In response to the NRC RAI request, Enclosure 2 contains a copy of the current Unit 1 TS section 3.3.1.1 and TS Bases. Some of these TS pages have been recently been changed as a result of NRC approval of various instrumentation TS changes, but do not show in Enclosure 2 since these recent TS Changes have not yet been implemented at the plant. These recent TS changes do not affect the OPRM TS-443 TS changes.

NRC RAI 7 Please describe in detail the current implementation status of the OPRM system for Browns Ferry Unit 1, including hardware and software modifications, operating experience for establishing the setpoints used to arm the proposed OPRM system, the detailed procedures to finalize system calibration and trip set-points based on the approach stated in NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," or based on a plant-specific data approach.

TVA Reply to RAI 7 The Unit 1 PRNM system including the OPRM functions is being installed by Design Change Notice (DCN) 51079. Implementation of this DCN is in progress and is expected to be field complete in October 2006. Post-modification testing will be accomplished by the performance of the OPRM SR Procedures (similar to those performed on Units 2 and 3) and performance of a Post-Modification Test Instruction (PMTI) that verifies proper response to a power supply loss. The post-modification testing is expected to be completed by November 2006.

Cycle-specific setpoint calculations are performed to determine the operating MCPR needed to protect the SLMCPR for the various OPRM amplitude setpoints. The results of this calculation are included in the cycle-specific core reload analysis (SRLR) report. The SRLR for Unit 1 Cycle 7 was transmitted to NRC on May 15, 2006 (Reference 3-1). The cycle-specific OPRM amplitude setpoint is chosen from these results such that it does not set the MCPR operating limit. The neutron monitoring system setpoint and scaling document is then revised as to implement the chosen amplitude setpoint and associated counts, and any required changes to the OPRM system are then implemented prior to unit startup.

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The OPRM trip function was not enabled during the first cycle of operation after initial installation on Units 2 and 3 to allow observation of system response during reactor operation.

Some false alarms were observed when operating in the OPRM enabled power range. Allowable tuning adjustments were made to the Units 2 and 3 OPRM systems to reduce the frequency of false alarm signals. Because of the similarities between Units 1, 2, and 3, the same allowable tuning adjustments will be made on the Unit 1 OPRM system. As noted in the TS-443 submittal, TVA proposes to restart Unit 1 with the OPRM system armed.

NRC RAI 8 NEDO-32465-A is a generic approved method for the DIVOM calculation, please provide a detailed description of the methodology for calculation of the plant-specific DIVOM correlation and identify the NRC-approved methodologies used to calculate the OPRM setpoints for TS 3.3.1.1.

TVA Reply to RAI 8 The OPRM amplitude setpoint calculation, as defined in NEDO-32465-A, is comprised of three components. These components and the underlying methodology are summarized below:

" Statistical Calculation of Peak Oscillation Magnitude Also known as the HCOM calculation. This calculation is not cycle dependent, but is dependent upon the OPRM configuration and trip system response characteristics. This calculation was performed for Units 2 and 3 using the GE OPRM code. The Unit 2 and 3 HCOMs remain valid for Unit 1 since the systems have the same configuration in regard to HCOM calculation inputs.

" MCPR Performance of the Hot Bundle Also known as DIVOM. Previously envisioned as a generic value, this has become a cycle-specific calculation in response to an earlier 10 CFR Part 21 concern. This is discussed in more detail later in this RAI response.

" Calculation of Pre-Oscillation MCPR This portion of the methodology ensures that the SLMCPR remains protected by ensuring that the correct pre-oscillation MCPR is used as a starting point. The reactor EI-II

simulator code, PANAC11, is used to establish the delta-CPR for a dual recirculation pump trip.

The generic regional mode DIVOM slope provided in NEDO-32465-A was determined to be potentially non-conservative as documented in the GE 10 CFR Part 21 notification, "Stability Reload Licensing Calculations Using Generic DIVOM Curve," MFN 01-046 dated August 31, 2001 (ML012490522). Closure of this issue using plant-specific DIVOM curves was endorsed by the NRC as documented in BWR Owners' Group (BWROG) letter BWROG-03047 dated September 30, 2003 (ML032751632).

The BWROG generated the "Plant-Specific Regional Mode DIVOM Procedure Guideline," GE-NE-0000-0028-9714-RI (Reference 3-1) to provide guidance to the fuel vendors in the calculation of DIVOM. Global Nuclear Fuel (GNF) is the fuel vendor for Unit 1 and is responsible for this calculation. GNF has incorporated the requirements of the BWROG guideline into their own internal technical procedures and these procedures were applied to the Unit 1 Cycle 7 reload licensing analyses.

NRC RAI 9 Provide detailed results of the system tests to support the accuracy and operability'of the current OPRM instrumentation.

Describe the data bases obtained during the shutdown and subsequent startup from refueling outage since the OPRM system was installed.

TVA Reply to RAI 9 Prior to installation, the Unit 1 PRNM system functionality was verified by Factory Acceptance Testing. Post-modification testing will be accomplished by the performance of surveillance procedures and a PMTI to verify proper response to a power supply loss.

The Unit 1 PRNM/OPRM SR Procedures are similar to the Unit 2 and 3 procedures. These SR procedures include the following:

0 Verification/change of entered setpoints, 0 Power supply calibration,

APRM and LPRM self-test,

Recorder calibrations,

Recirculation flow loop calibrations,

Average Power Range Monitor/LPRM/OPRM functional tests, 0 RBM calibrations and functional tests, and El-12

0 Voter logic functional tests.

The plant Integrated Computer System (ICS) monitors the OPRM parameters with a sampling rate of approximately once per second. This data is available real time on an ICS display with approximately 6 minutes of data history available on the display. This data is also archived in an online data base such that the previous three years of OPRM data may be readily retrieved.

Although Unit 1 OPRM operating data cannot be obtained until Unit 1 returns to service, Units 2 and Unit 3 have been now been operating with the OPRM Option III system enabled since the Springs of 1999 and 2000, respectively. Thus, BFN has ample operating experience with the OPRM system in service and OPRM operating characteristics at power are well understood.

NRC RAI 10 According to the Boiling Water Reactors Owners' Group (BWROG) letter, BWROG-03049 dated September 30, 2003, "Utility Commitment to NRC for OPRM Operability at Option III Plants," a plant-specific DIVOM curve is recommended. Please identify any plant-specific differences from the generic values specified in NEDO-32465-A such as Period Based Detection Algorithm (PBDA) period confirmation setpoints in Table 3-1, PBDA trip setpoints in Table 3-2, and generic DIVOM curve slope. Also, provide plant-specific values for OPRM scram setpoints and the DIVOM correlation for the next cycle.

TVA Reply to RAI 10 Table 3-1 of NEDO-32465-A (Reference 3-2)identifies ranges of 100 to 300 msec for period tolerance (ý) and 1 to 2.5 Hz for the cutoff frequency (Fc). As stated in the response to RAI 1, the BFN values are 100 msec and 1 Hz, which are within the specified range.

Table 3-2 of NEDO-32465-A identifies the relationship between confirmation counts (Np) and Amplitude Setpoint (Sp) for PBDA.

This relationship is used at BFN.

As noted in the response to RAI 8, Unit 1 will be using a plant-specific DIVOM slope. The Unit 1 Cycle 7 DIVOM calculation resulted in a slope of 0.49. The OPRM amplitude setpoint (Sp) versus required OLMCPR is provided in Table 10-1 El-13

below. Both the DIVOM slope and resulting OPRM setpoints were previously provided in section 15.2 of the Unit 1 SRLR (Reference 3-1).

Table 10-1 Unit 1 Cycle 7 OPRM Amplitude Setpoint OPRM OLMCPR OLMCPR(2PT)

Amplitude (Steady Set point State) 1.05 1.1852 1.1067 1.06 1.2053 1.1254 1.07 1.2260 1.1447 1.08 1.2475 1.1648 1.09 1.2697 1.1855 1.10 1.2927 1.2070 1.11 1.3155 1.2283 1.12 1.3391 1.2503 1.13 1.3635 1.2731 1.14 1.3889 1.2968 1.15 1.4152 1.3214 Acceptance Off-rated Rated Power Criteria OLMCPR OLMCPR as

@45% flow described in SRLR Section 11 NRC RAI 11 Please describe the alternate method to detect and suppress thermal hydraulic instability oscillations as stated in Condition I.1 of Actions under LCO 3.3.1.1.

TVA Reply to RAI 11 BFN Unit 1 will implement the Backup Stability Protection (BSP) regions defined in 0G02-0119-260, "Backup Stability Protection (BSP) for Inoperable Option III Solution," dated July 17, 2002.

This solution is based upon scram and controlled entry regions on the power/flow operating map. Each region is defined on a cycle-specific basis based upon calculated decay ratios. The codes and acceptance criteria used are fuel vendor specific.

For Unit 1, the ODYSY code is used for the calculation of decay ratios based upon statepoint and neutronic data is from PANACII and TGLBLA06.

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The alternate detect and suppress method is enforced by unit specific SR procedure SR-3.3.1.1.I. This procedure applies to Condition I of TS Limiting Condition for Operation 3.3.1.1 (OPRM inoperable) and contains the power/flow maps with the required regions as well as necessary operator actions. These actions include the requirement to scram if the reactor enters Region 1. Additionally, the power/flow map is displayed by the plant process computer for operator use. This is similar to the approach currently being used on BFN Units 2 and 3.

NRC RAI 12 Normally specific setpoint limits for OPRM are specified in the core operating limit report. Please justify that there are no setpoints for the trip function specified in proposed Surveillance for OPRM or note to the Table 3.3.1.1-1 under Function 2.f.

TVA Reply to RAI 12 At BFN, the OPRM setpoints are maintained in Appendix N of the Updated Final Safety Analysis Report (UFSAR), which contains the cycle-specific reload analysis for each BFN unit. Changes to the UFSAR (and hence OPRM setpoints) are subject to evaluation under 10 CFR 50.59. Therefore, BFN considers maintenance of the OPRM setpoints in the UFSAR to be equivalent to maintenance in the Core Operating Limits Report and is subject to a similar change control process.

The current BFN Units 2 and 3 TS, and the proposed TS changes in TS-443 do not list the OPRM setpoints in Table 3.3.1.1-1 under Function 2.f. This is consistent with the model TS shown in Table 3.3.1.1-1 of NEDC-32410PA Supplement 1 (Reference 4-1), which NRC approved on April 15, 1997. The setpoints tend to be cycle-specific, which would result in a excessive need for license amendments if they were located in TS.

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ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 1 TS-443 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING OSCILLATION POWER RANGE MONITOR (OPRM)

COPY OF BFN UNIT 1 TS SECTION 3.3.1.1 AND BASES

RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1.

ACTIONS NOTE-Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channels inoperable.

OR A.2 Place associated trip 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system in trip.

B. One or more Functions B.1 Place channel in one trip 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with one or more required system in trip.

channels inoperable in both trip systems. OR B.2 Place one trip system in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months trip.

C. One or more Functions C.1 Restore RPS trip 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability capability.

not maintained.

(continued)

BFN-UNIT 1 3.3-1 Amendment No. 234

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, or Table 3.3.1.1-1 for the C .not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action D.1 and POWER to < 30% RTP.

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully Immediately Action D.1 and insert all insertable referenced in control rods in core cells Table 3.3.1.1-1. containing one or more fuel assemblies.

BFN-UNIT 1 3.3-2 Amendment No. 234

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS NOTES

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.3.1.1.2 NOTE Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER Ž 25% RTP.

Verify the absolute difference between the 7 days average power range monitor (APRM) channels and the calculated power is

-<2% RTP plus any gain adjustment required by LCO 3.2.4, "Average Power Range Monitor (APRM) Setpoints" while operating at

Ž25% RTP.

SR 3.3.1.1.3 NOTE Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 7 days (continued)

BFN-UNIT 1 3.3-3 Amendment No. 234

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position SR 3.3.1.1.6 -NOTE Only required to be met during entry into MODE 2 from MODE 1.

Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.7 Calibrate the local power range monitors. 1000 MWD/T average core exposure SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.9 NOTES-

1. Neutron detectors are excluded.

2. For Functions 1 and 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL CALIBRATION. 92 days (continued)

BFN-UNIT 1 3.3-4 Amendment No. 234

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.10 Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.1.11 Adjust the channel to conform to a calibrated 18 months flow signal.

SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 18 months SR 3.3.1.1.13 Perform CHANNEL CALIBRATION. 18 months SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL 18 months TEST.

SR 3.3.1.1.15 Verify Turbine Stop Valve-Closure and 18 months Turbine Control Valve Fast Closure, Trip Oil Pressure -Low Functions are not bypassed when THERMAL POWER is Ž_30% RTP.

BFN-UNIT 1 3.3-5 Amendment No. 234

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 FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 s120/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1 5 120/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.9 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5 (a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors

a. Neutron Flux - High, 2 2 G SR 3.3.1.1.1 < 15% RTP Setdown SR 3.3.1.1.3 SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.9 SR 3.3.1.1.14

b. Flow Biased Simulated 1 2 F SR 3.3.1.1.1 <0.58W Thermal Power - High SR 3.3.1.1.2 + 62% RTP SR 3.3.1.1.7 and <120%

SR SR 3.3.1.1.8 3.3.1.1.9 RTP(b) I SR 3.3.1.1.11 SR 3.3.1.1.14

c. Neutron Flux - High I 2 F SR 3.3.1.1.1 < 120% RTP SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.9 SR 3.3.1.1.14 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) [0.58 W + 62% - 0.58 A W] RTP when reset for single loop operation per LCO 3.4.1, *Recirculatlon Loops Operating.

BFN-UNIT 1 3.3-6 Amendment No. 236 December 23, 1998

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 FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

2. Average Power Range Monitors (continued)

d. Downscale 1 2 F SR 3.3.1.1.7 2:3% RTP SR 3.3.1.1.8 SR 3.3.1.1.14

e. Inop 1,2 2 G SR 3.3.1.1.7 NA SR 3.3.1.1.8 SR 3.3.1.1.14

3. Reactor Vessel Steam Dome 1,2 2 G SR 3.3.1.1.1 S 105 5 psig Pressure - High SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14

4. Reactor Vessel Water Level - 1,2 2 G SR 3.3.1.1.1 > 538 inches Low, Level 3 SR 3.3.1.1.8 above vessel SR 3.3.1.1.13 zero SR 3.3.1.1.14

5. Main Steam Isolation Valve - 1 8 F SR 3.3.1.1.8 < 10% closed Closure SR 3.3.1.1.13 SR 3.3.1.1.14

6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.8 :5 2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14

7. Scram Discharge Volume Water Level - High

a. Resistance Temperature 1.2 2 G SR 3.3.1.1.8 < 50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.8 < 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

b. Float Switch 1,2 2 G SR 3.3.1.1.8 :5 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 5 (a) 2 H SR 3.3.1.1.8 < 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

BFN-UNIT 1 3.3-7 Amendment No. 234

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

8. Turbine Stop Valve - Closure >:30% RTP 4 E SR 3.3.1.1.8 : 10% closed SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15

9. Turbine Control Valve Fast a 30% RTP 2 E SR 3.3.1.1.8 a 550 psig Closure, Trip Oil Pressure - SR 3.3.1.1.13 Low SR 3.3.1.1.14 SR 3.3.1.1.15

10. Reactor Mode Switch - 1,2 1 G SR 3.3.1.1.12 NA Shutdown Position SR 3.3.1.1.14 5 (a) 1 H SR 3.3.1.1.12 NA SR 3.3.1.1.14

11. Manual Scram 1,2 1 G SR 3.3.1.1.8 NA SR 3.3.1.1.14 5 (a) I H SR 3.3.1.1.8 NA SR 3.3.1.1.14

12. RPS Channel Test Switches 1,2 2 G SR 3.3.1.1.4 NA 5 (a) 2 H SR 3.3.1.1.4 NA (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

BFN-UNIT 1 3.3-8 Amendment No. 234

RPS Instrumentation B 3.3.1.1 B 3.3 INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) Instrumentation BASES BACKGROUND The RPS initiates a reactor scram when one or more monitored parameters exceed their specified limits, to preserve the integrity of the fuel cladding and the Reactor Coolant System (RCS) and minimize the energy that must be absorbed following a loss of coolant accident (LOCA). This can be accomplished either automatically or manually.

The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. The LSSS are defined in this Specification as the Allowable Values, which, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits, including Safety Limits (SLs) during Design Basis Accidents (DBAs).

The RPS, as described in the FSAR, Section 7.2 (Ref. 1),

includes sensors, relays, bypass dircuits, and switches that are necessary to cause initiation of a reactor scram. Functional diversity is provided by monitoring a wide range of dependent and independent parameters. The input parameters to the scram logic are from instrumentation that monitors reactor vessel water level, reactor vessel pressure, neutron flux, main steam line isolation valve position, turbine control valve (TCV)

(continued)

BFN-UNIT 1 B 3.3-1 Revision 0

RPS Instrumentation B 3.3.1.1 BASES BACKGROUND fast closure trip oil pressure (indicated by TCV low hydraulic (continued) pressure), turbine stop valve (TSV) position, drywell pressure, and scram discharge volume (SDV) water level, as well as reactor mode switch in shutdown position, manual, and RPS channel test switch scram signals. There are at least four redundant sensor input signals from each of these parameters (with the exception of the reactor mode switch in shutdown, manual, and RPS channel test switch scram signals). Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay deenergizes, which then outputs an RPS trip signal to the trip logic.

The RPS is comprised of two independent trip systems (A and B) with two logic channels in each trip system (logic channels Al and A2, B1 and 82) as shown in Reference 1. The outputs of the logic channels in a trip system are combined in a one-out-of-two logic so that either channel can trip the associated trip system. The tripping of both trip systems will produce a reactor scram. This logic arrangement is referred to as a one-out-of-two taken twice logic. Each trip system can be reset by use of a reset switch. If a full scram occurs (both trip systems trip), a relay prevents reset of the trip systems for 10 seconds after the full scram signal is received. This 10 second delay on reset ensures that the scram function will be completed.

(continued)

BFN-UNIT 1 B 3.3-2 Revision 0

RPS Instrumentation B 3.3.1.1 BASES BACKGROUND Two scram pilot valves are located in the hydraulic control unit (continued) for each control rod drive (CRD). Each scram pilot valve is solenoid operated, with the solenoids normally energized. The scram pilot valves control the air supply to the scram inlet and outlet valves for the associated CRD. When either scram pilot valve solenoid is energized, air pressure holds the scram valves closed and, therefore, both scram pilot valve solenoids must be de-energized to cause a control rod to scram. The scram valves control the supply and discharge paths for the CRD water during a scram. One of the scram pilot valve solenoids for each CRD is controlled by trip system A, and the other solenoid is controlled by trip system B. Any trip of trip system A in conjunction with any trip in trip system B results in de-energizing both solenoids, air bleeding off, scram valves opening, and control rod scram.

The backup scram valves, which energize on a full scram signal to depressurize the scram air header, are also controlled by the RPS. Additionally, the RPS System controls the SDV vent and drain valves such that when both trip systems trip, the SDV vent and drain valves close to isolate the SDV.

APPLICABLE The actions of the RPS are assumed in the safety analyses of SAFETY ANALYSES, References 1, 2, and 3. The RPS initiates a reactor scram LCO, and when monitored parameter values exceed the Allowable APPLICABILITY Values, specified by the setpoint methodology and listed in Table 3.3.1.1-1 to preserve the integrity of the fuel cladding, the reactor coolant pressure boundary (RCPB), and the containment by minimizing the energy that must be absorbed following a LOCA.

-(continued)

BFN-UNIT I B 3.3-3 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE RPS instrumentation satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES, Statement (Ref. 10). Functions not specifically credited in the LCO, and accident analysis are retained for the overall redundancy and APPLICABILITY diversity of the RPS as required by the NRC approved licensing (continued) basis.

The OPERABILITY of the RPS is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.1.1-1. Each Function must have a required number of OPERABLE channels per RPS trip system, with their setpoints within the specified Allowable Value, where appropriate. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint).

Allowable Values are specified for each RPS Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.

Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g.,

trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.

(continued)

BFN-UNIT 1 B 3.3-4 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE The trip setpoints are then determined accounting for the SAFETY ANALYSES, remaining instrument errors (e.g., drift). The trip setpoints LCO, and derived in this manner provide adequate protection because APPLICABILITY instrumentation uncertainties, process effects, calibration (continued) tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in the Background section, are not addressed by this LCO.

The individual Functions are required to be OPERABLE in the MODES or other specified conditions in the Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required in each MODE to provide primary and diverse initiation signals.

The only MODES specified in Table 3.3.1.1-1 are MODES 1 (which encompasses > 30% RTP) and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. No RPS Function is required in MODES 3 and 4 since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LCO 3.3.2.1) does not allow any control rod to be withdrawn.

In MODE 5, control rods withdrawn from a core cell containing no fuel assemblies do not affect the reactivity of the core and, therefore, are not required to have the capability to scram.

Provided all other control rods remain inserted, no RPS function is required. In this condition, the required SDM (LCO 3.1.1) and refuel position one-rod-out interlock (LCO 3.9.2) ensure that no event requiring RPS will occur.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

(continued)

BFN-UNIT 1 B 3.3-5 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE Intermediate Rancqe Monitor (IRM)

SAFETY ANALYSES, LCO, and l.a. Intermediate Ran-qe Monitor Neutron Flux - High APPLICABILITY (continued) The IRMs monitor neutron flux levels from the upper range of the source range monitor (SRM) to the lower range of the average power range monitors (APRMs). The IRMs are capable of generating trip signals that can be used to prevent fuel damage resulting from abnormal operating transients in the intermediate power range. In this power range, the most significant source of reactivity change is due to control rod withdrawal. The IRM mitigates control rod withdrawal error events and is diverse from the rod worth minimizer (RWM),

which monitors and controls the movement of control rods at low power. The RWM prevents the withdrawal of an out of sequence control rod during startup that could result in an unacceptable neutron flux excursion (Ref. 2). The IRM provides mitigation of the neutron flux excursion. To demonstrate the capability of the IRM System to mitigate control rod withdrawal events, generic analyses have been performed (Ref. 3) to evaluate the consequences of control rod withdrawal events during startup that are mitigated only by the IRM. This analysis, which assumes that one IRM channel in each trip system is bypassed, demonstrates that the IRMs provide protection against local control rod withdrawal errors and results in peak fuel energy depositions below the 170 cal/gm fuel failure threshold criterion.

The IRMs are also capable of limiting other reactivity excursions during startup, such as cold water injection events, although no credit is specifically assumed.

(continued)

BFN-UNIT 1 B 3.3-6 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE l.a. Intermediate Range Monitor Neutron Flux - High SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY The IRM System is divided into two groups of IRM channels, with four IRM channels inputting to each trip system. The analysis of Reference 3 assumes that one channel in each trip system is bypassed. Therefore, six channels with three channels in each trip system are required for IRM OPERABILITY to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This trip is active in each of the 10 ranges of the IRM, which must be selected by the operator to maintain the neutron flux within the monitored level of an IRM range.

The analysis of Reference 3 has adequate conservatism to permit an IRM Allowable Value of 120 divisions of a 125 division scale.

The Intermediate Range Monitor Neutron Flux - High Function must be OPERABLE during MODE 2 when control rods may be withdrawn and the potential for criticality exists. In MODE 5, when a cell with fuel has its control rod withdrawn, the IRMs provide monitoring for and protection against unexpected reactivity excursions. In MODE 1, the APRM System and the RBM provide protection against control rod withdrawal error events and the IRMs are not required.

(continued)

BFN-UNIT 1 B 3.3-7 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 1.b. Intermediate Ran-qe Monitor - Inop SAFETY ANALYSES, LCO, and This trip signal provides assurance that a minimum number of APPLICABILITY IRMs are OPERABLE. Anytime an IRM mode switch is moved (continued) to any position other than "Operate," the detector voltage drops below a preset level, or when a module is not plugged in, an inoperative trip signal will be received by the RPS unless the IRM is bypassed. Since only one IRM in each trip system may be bypassed, only one IRM in each RPS trip system may be inoperable without resulting in an RPS trip signal.

This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

Six channels of Intermediate Range Monitor - Inop with three channels in each trip system are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

Since this Function is not assumed in the safety analysis, there is no Allowable Value for this Function.

This Function is required to be OPERABLE when the Intermediate Range Monitor Neutron Flux - High Function is required.

(continued)

BFN-UNIT 1 B 3.3-8 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE Average Power Ran-ge Monitor SAFETY ANALYSES, LCO, and 2.a. Average Power Range Monitor Neutron Flux - High, APPLICABILITY Setdown (continued)

The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes.

The APRM channels average these LPRM signals to provide a continuous indication of average reactor power from a few percent to greater than RTP. 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 Function will provide a secondary scram to the Intermediate Range Monitor Neutron 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 Flux - High, Setdown Function will provide the primary trip signal for a corewide increase in power.

No specific safety analyses take direct credit for the Average Power Range Monitor Neutron Flux - High, Setdown Function.

However, this Function indirectly ensures that before the reactor mode switch is placed in the run position, reactor power does not exceed 25% 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 with THERMAL POWER < 25% RTP.

(continued)

BFN-UNIT 1 B 3.3-9 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Averaqe Power Range Monitor Neutron Flux - High, SAFETY ANALYSES, Setdown (continued)

LCO, and APPLICABILITY The APRM System is divided into two groups of channels with three APRM channel inputs to each trip system. The system is designed to allow one channel in each trip system to be bypassed. Any one APRM channel in a trip system can cause the associated trip system to trip. Four channels of Average Power Range Monitor Neutron Flux - High, Setdown with two channels in each trip system are required to be OPERABLE to ensure that no single failure will preclude a scram from this Function on a valid signal. In addition, to provide adequate coverage of the entire core, at least 14 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located.

The Allowable Value is based on preventing significant increases in power when THERMAL POWER is < 25% RTP.

The Average Power Range Monitor Neutron Flux - High, Setdown Function must be OPERABLE during MODE 2 when control rods may be withdrawn since the potential for criticality exists.

In MODE 1, the Average Power Range Monitor Neutron Flux -

High Function provides protection against reactivity transients and the RWM and rod block monitor protect against control rod withdrawal error events.

(continued)

BFN-UNIT 1 B 3.3-10 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES, Thermal Power - High LCO, and APPLICABILITY The Average Power Range Monitor Flow Biased Simulated (continued) 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 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 recirculation drive flow (i.e., at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed control rod pattern) but is clamped at an upper limit that is always lower than or equal to the Average Power Range Monitor Fixed Neutron Flux - High Function Allowable Value. The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function provides protection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating event) and protects the fuel cladding integrity by ensuring that the MCPR SL is not exceeded. During these events, the THERMAL POWER increase does not significantly lag the neutron flux response and, because of a lower trip setpoint, will initiate a scram before the high neutron flux scram.

For rapid neutron flux increase events, the THERMAL POWER lags the neutron flux and the Average Power Range Monitor Fixed Neutron Flux - High Function will provide a scram signal before the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function setpoint is exceeded.

(continued)

BFN-UNIT 1 B 3.3-11 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES, Thermal Power - High (continued)

LCO, and APPLICABILITY The APRM System is divided into two groups of channels with three APRM channel inputs to each trip system. The system is designed to allow one channel in each trip system to be bypassed. Any one APRM channel in a trip system can cause the associated trip system to trip. Four channels of Average Power Range Monitor Flow Biased Simulated Thermal Power -

High with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. In addition, to provide adequate coverage of the entire core, at least 14 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located. Each APRM channel receives a total drive flow signal representative of total core flow. The total drive flow signals are generated by two flow units, one of which supplies signals to the trip system A APRMs, while the other one supplies signals to the trip system B APRMs. Each flow unit signal is provided by summing up the flow signals from the two recirculation loops. Each required Average Power Range Monitor Flow Biased Simulated Thermal Power - High channel requires an input from its associated OPERABLE flow unit.

The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function for the mitigation of the loss of feedwater heating event. The THERMAL POWER time constant of < 7 seconds is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER. The term "W" in the equation for determining the Allowable Value is defined as total recirculation flow in percent of rated.

(continued)

BFN-UNIT 1 B 3.3-12 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES, Thermal Power - High (continued)

LCO, and APPLICABILITY The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.

2.c. Averaae Power Ranae Monitor Fixed Neutron Flux - Hiah The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases. The Average Power Range Monitor Fixed Neutron Flux - High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure.

For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Fixed Neutron Flux - High Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety/relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Fixed Neutron Flux - High Function to terminate the CRDA.

(continued)

BFN-UNIT 1 B 3.3-13 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.c. Average Power Range Monitor Fixed Neutron Flux - High SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY The APRM System is divided into two groups of channels with three APRM channels inputting to each trip system. The system is designed to allow one channel in each trip system to be bypassed. Any one APRM channel in a trip system can cause the associated trip system to trip. Four channels of Average Power Range Monitor Fixed Neutron Flux - High with two channels in each trip system arranged in a one-out-of- two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. In addition, to provide adequate coverage of the entire core, at least 14 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located.

The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.

The Average Power Range Monitor Fixed Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded.

Although the Average Power Range Monitor Fixed Neutron Flux

- High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High, Setdown Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Fixed Neutron Flux - High Function is not required in MODE 2.

(continued)

BFN-UNIT 1 B 3.3-14 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.d. Averaae Power Ranqe Monitor - Downscale SAFETY ANALYSES, LCO, and This signal ensures that there is adequate Neutron Monitoring APPLICABILITY System protection if the reactor mode switch is placed in the (continued) run position prior to the APRMs coming on scale. With the reactor mode switch in run, an APRM downscale signal coincident with an associated Intermediate Range Monitor Neutron Flux - High or Inop signal generates a trip signal. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The APRM System is divided into two groups of channels with three inputs into each trip system. The system is designed to allow one channel in each trip system to be bypassed. Four channels of Average Power Range Monitor - Downscale with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single failure will preclude a scram from this Function on a valid signal. The Intermediate Range Monitor Neutron Flux - High and Inop Functions are also part of the OPERABILITY of the Average Power Range Monitor -. Downscale Function (i.e., if either of these IRM Functions cannot send a signal to the Average Power Range Monitor - Downscale Function, the associated Average Power Range Monitor - Downscale channel is considered inoperable).

The Allowable Value is based upon ensuring that the APRMs are in the linear scale range when transfers are made between APRMs and IRMs.

This Function is required to be OPERABLE in MODE 1 since this is when the APRMs are the primary indicators of reactor power.

(continued)

BFN-UNIT 1 B 3.3-15 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.e. Averaae Power Ranae Monitor - InoD SAFETY ANALYSES, LCO, and This signal provides assurance that a minimum number of APPLICABILITY APRMs are OPERABLE. Anytime an APRM mode switch is (continued) moved to any position other than "Operate," an APRM module is unplugged, the electronic operating voltage is low, or the APRM has too few LPRM inputs (< 14), an inoperative trip signal will be received by the RPS, unless the APRM is bypassed. Since only one APRM in each trip system may be bypassed, only one APRM in each trip system may be inoperable without resulting in an RPS trip signal. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

Four channels of Average Power Range Monitor - Inop with two channels in each trip system are required to be OPERABLE to ensure that no single failure will preclude a scram from this Function on a valid signal.

There is no Allowable Value for this Function.

This Function is required to be OPERABLE in the MODES

.where the APRM Functions are required.

(continued)

BFN-UNIT 1 B 3.3-16 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 3. Reactor Vessel Steam Dome Pressure - Hi-qh SAFETY ANALYSES, (PIS-3-22AA, PIS-3-22BB, PIS-3-22C, and PIS-3-22D)

LCO, and APPLICABILITY An increase in the RPV pressure during reactor operation (continued) 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 result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 4, reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Fixed Neutron Flux - High signal, not the Reactor Vessel Steam Dome Pressure - High signal), along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.

High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure - High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.

Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.

(continued)

BFN-UNIT 1 B 3.3-17 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 4. Reactor Vessel Water Level - Low, Level 3 SAFETY ANALYSES, (LIS-3-203A, LIS-3-203B, LIS-3-203C, and LIS-3-203D)

LCO, and APPLICABILITY Low RPV water level indicates the capability to cool the fuel (continued) may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level - Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 6). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

Four channels of Reactor Vessel Water Level - Low, Level 3 Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

The Reactor Vessel Water Level - Low, Level 3 Allowable Value is selected to ensure that (a) during normal operation the steam dryer skirt is not uncovered (this protects available recirculation pump net positive suction head (NPSH) from significant carryunder), and (b) for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water - Low Low Low, Level 1 will not be required.

(continued)

BFN-UNIT 1 B 3.3-18 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 4. Reactor Vessel Water Level - Low, Level 3 SAFETY ANALYSES, (LIS-3-203A, LIS-3-203B, LIS-3-203C, and LIS-3-203D)

LCO, and (continued)

APPLICABILITY The Function is required in MODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. ECCS initiations at Reactor Vessel Water Level - Low Low, Level 2 and Low Low Low, Level 1 provide sufficient protection for level transients in all other MODES.

5. Main Steam Isolation Valve - Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient.

However, for the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Fixed Neutron Flux - High Function, along with the S/RVs, limits the peak RPV pressure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis. Additionally, MSIV closure is assumed in the transients analyzed in Reference 7 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow).

The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

(continued)

BFN-UNIT 1 B 3.3-19 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 5. Main Steam Isolation Valve - Closure (continued)

SAFETY ANALYSES, LCO, and MSIV closure signals are initiated from position switches APPLICABILITY located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve -

Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve - Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur.

The Main Steam Isolation Valve - Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.

Sixteen channels of the Main Steam Isolation Valve - Closure Function, with eight channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This Function is only required in MODE 1 since, with the MSIVs open and the heat generation rate high, a pressurization transient can occur if the MSIVs close. In MODE 2, the heat generation rate is low enough so that the other diverse RPS functions provide sufficient protection.

(continued)

BFN-UNIT 1 B 3.3-20 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 6. Drywell Pressure - High SAFETY ANALYSES, (PIS-64-56A, PIS-64-56B, PIS-64-56C, and PIS-64-56D)

LCO, and APPLICABILITY High pressure in the drywell could indicate a break in the (continued) RCPB. A reactor scram is initiated to minimize the possibility of fuel damage and to reduce the amount of energy being added to the coolant and the drywell. The Drywell Pressure - High Function is a secondary scram signal to Reactor Vessel Water Level - Low, Level 3 for LOCA events inside the drywell.

However, no credit is taken for a scram initiated from this Function for any of the DBAs analyzed in the FSAR. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and indicative of a LOCA inside primary containment.

Four channels of Drywell Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS, resulting in the limiting transients and accidents.

(continued)

BFN-UNIT I B 3.3-21 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 7a, 7b. Scram Discharge Volume Water Level - Hiqh SAFETY ANALYSES, (LS-85-45A, LS-85-45B, LS-85-45C, LS-85-45D, LCO, and LS-85-45E, LS-85-45F, LS-85-45G, and LS-85-45H)

APPLICABILITY (continued) The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered.

Therefore, a reactor scram is initiated while the remaining free volume is still sufficient to accommodate the water from a full core scram. The two types of Scram Discharge Volume Water Level - High Functions are an input to the RPS logic. No credit is taken for a scram initiated from these Functions for any of the design basis accidents or transients analyzed in the FSAR.

However, they are retained to ensure the RPS remains OPERABLE.

SDV water level is measured by two diverse methods. The level in each of the two SDVs is measured by two float type level switches and two thermal probes for a total of eight level signals. The outputs of these devices are arranged so that there is a signal from a level switch and a thermal probe to each RPS logic channel. The level measurement instrumentation satisfies the recommendations of Reference 8.

The Allowable Value is chosen low enough to ensure that there is sufficient volume in the SDV to accommodate the water from a full scram.

(continued)

BFN-UNIT 1 B 3.3-22 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 7a, 7b. Scram Discharge Volume Water Level - High SAFETY ANALYSES, (LS-85-45A, LS-85-45B, LS-85-45C, LS-85-45D, LCO, and LS-85-45E, LS-85-45F, LS-85-45G, and LS-85-45H)

APPLICABILITY (continued)

Four channels of each type of Scram Discharge Volume Water Level - High Function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal. These Functions are required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.

8. Turbine Stop Valve - Closure Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated at the start of TSV closure in anticipation of the transients that would result from the closure of these valves.

The Turbine Stop Valve - Closure Function is the primary scram signal for the turbine trip event analyzed in Reference 7.

For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded.

(continued)

BFN-UNIT I B, 3.3-23 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 8. Turbine Stop Valve - Closure (continued)

SAFETY ANALYSES, LCO, and Turbine Stop Valve - Closure signals are initiated from position APPLICABILITY switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram. This Function must be enabled at THERMAL POWER _>30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function.

The Turbine Stop Valve - Closure Allowable Value is selected to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.

Eight channels of Turbine Stop Valve - Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function if any three TSVs should close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is > 30% RTP.

This Function is not required when THERMAL POWER is

< 30% RTP since the Reactor Vessel Steam Dome Pressure -

High and the Average Power Range Monitor Fixed Neutron Flux

- High Functions are adequate to maintain the necessary safety margins.

(continued)

BFN-UNIT 1 B 3.3-24 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 9. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low SAFETY ANALYSES, (PS-47-142, PS-47-144, PS-47-146, and PS-47-148)

LCO, and APPLICABILITY Fast closure of the TCVs results in the loss of a heat sink that (continued) produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Function is the primary scram signal for the generator load rejection event analyzed in Reference 7. For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL is not exceeded.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve. One pressure switch is associated with each control valve, and the signal from each switch is assigned to a separate RPS logic channel. This Function must be enabled at THERMAL POWER _>30% RTP.

This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function.

The Turbine Control Valve Fast Closure, Trip Oil Pressure -

Low Allowable Value is selected high enough to detect imminent TCV fast closure.

(continued)

BFN-UNIT 1 B 3.3-25 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 9. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low SAFETY ANALYSES, (PS-47-142, PS-47-144, PS-47-146, and PS-47-148)

LCO, and (continued)

APPLICABILITY Four channels of Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required, consistent with the analysis assumptions, whenever THERMAL POWER is > 30% RTP. This Function is not required when THERMAL POWER is < 30% RTP, since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Fixed Neutron Flux - High Functions are adequate to maintain the necessary safety margins.

10. Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, directly to the scram pilot solenoid power circuits. These manual scram logic channels are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The reactor mode switch is a single switch with four channels, each of which provides input into one of the RPS logic channels.

(continued)

BFN-UNIT 1 B 3.3-26 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 10. Reactor Mode Switch - Shutdown Position (continued)

SAFETY ANALYSES, LCO, and There is no Allowable Value for this Function, since the APPLICABILITY channels are mechanically actuated based solely on reactor mode switch position.

Two channels of Reactor Mode Switch - Shutdown Position Function, with one channel in each trip system, are available and required to be OPERABLE. The Reactor Mode Switch -

Shutdown Position Function is required to be OPERABLE in MODES 1 and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.

11. Manual Scram The Manual Scram push button channels provide signals, via the manual scram logic channels, directly to the scram pilot solenoid power circuits. These manual scram logic channels are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

There is one Manual Scram push button channel for each of the two RPS manual scram logic channels. In order to cause a scram it is necessary that each channel in both manual scram trip systems be actuated.

There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons.

(continued)

BFN-UNIT 1 B 3.3-27 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 11. Manual Scram (continued)

SAFETY ANALYSES, LCO, and Two channels of Manual Scram with one channel in each APPLICABILITY manual scram trip system are available and required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.

12. RPS Channel Test Switches There are four RPS Channel Test Switches, one associated with each of the four automatic scram logic channels (Al, A2, B1, and B2). These keylock switches allow the operator to test the OPERABILITY of each individual logic channel without the necessity of using a scram function trip. When the RPS Channel Test Switch is placed in test, the associated scram logic channel is deenergized and OPERABILITY of the channel's scram contactors can be confirmed. The RPS Channel Test Switches are not specifically credited in the accident analysis. However, because the Manual Scram Function at Browns Ferry Nuclear Plant is not configured the same as the generic model in Reference 9, the RPS Channel Test Switches are included in the analysis in Reference 11.

Reference 11 concludes that the Surveillance Frequency extensions for RPS functions, described in Reference 9, are not affected by the difference in configuration since each automatic RPS channel has a test switch which is functionally the same as the manual scram switches in the generic model. Weekly testing of scram contactors is credited in Reference 9 with supporting the Surveillance Frequency extension of the RPS functions.

(continued)

BFN-UNIT 1 B 3.3-28 Revision 0

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 12. RPS Channel Test Switches (continued)

SAFETY ANALYSES, LCO, and There is no Allowable Value for this Function since the APPLICABILITY channels are mechanically actuated solely on the position of the switches.

Four channels of the RPS Channel Test Switch Function with two channels in each trip system arranged in a one-out-of-two logic are available and required to be OPERABLE. The function is required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.

ACTIONS A Note has been provided to modify the ACTIONS related to RPS instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel.

(continued)

BFN-UNIT 1 B 3.3-29 Revision 0

RPS Instrumentation B 3.3.1.1 BASES ACTIONS A.1 and A.2 (continued)

Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months has been shown to be acceptable (Ref. 9) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken.

B.1 and B.2 Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip system. In this condition, provided at least one channel per trip system is OPERABLE, the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system.

(continued)

BFN-UNIT 1 B 3.3-30 Revision 0

RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued)

Required Actions B.1 and B.2 limit the time the RPS scram logic, for any Function, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function). The reduced reliability of this logic arrangement was not evaluated in Reference 9 for the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time. Within the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system.

Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in Reference 9, which justified a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowable out of service time as presented in Condition A. The trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in). If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip.

(continued)

BFN-UNIT 1 B 3.3-31 Revision 0

RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued)

The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.

Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip system in trip would result in a scram or RPT), Condition D must be entered and its Required Action taken.

C..1 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. For the typical Function with one-out-of-two taken twice logic and the IRM and APRM Functions, this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip). For Function 5 (Main Steam Isolation Valve -

Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or in trip (or the associated trip system in trip).

(continued)

BFN-UNIT 1 B 3.3-32 Revision 0

RPS Instrumentation B 3.3.1.1 BASES ACTIONS C. 1 (continued)

For Function 8 (Turbine Stop Valve - Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip).

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

D._1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition.

E.1, F.1, and G.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 Action E.1 is consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)."

(continued)

BFN-UNIT 1 B 3.3-33 Revision 0

RPS Instrumentation B 3.3.1.1 BASES ACTIONS H.1 (continued)

Ifthe 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.

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the associated Function maintains RPS trip capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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. 3) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the RPS will trip when necessary.

(continued)

BFN-UNIT 1 B 3.3-34 Revision 0

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. Ifa channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

(continued)

BFN-UNIT 1 B 3.3-35 Revision 0

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.2 REQUIREMENTS (continued) To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. LCO 3.2.4, "Average Power Range Monitor (APRM) Gain and Setpoints," allows the APRMs to be reading greater than actual THERMAL POWER to compensate for localized power peaking. When this adjustment is made, the requirement for the APRMs to indicate within 2% RTP of calculated power is modified to require the APRMs to indicate within 2% RTP of calculated MFLPD. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading, between performances of SR 3.3.1.1.7.

A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at > 25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when

< 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At Ž 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

(continued)

BFN-UNIT 1 B 3.3-36 Revision 0

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.3 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

As noted, SR 3.3.1.1.3 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM and APRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links.

This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9).

SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A Frequency of 7 days provides an acceptable level of system average availability over the Frequency and is based on the reliability analysis of Reference 9. (The RPS Channel Test Switch Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions' Frequencies.)

(continued)

BFN-UNIT 1 B 3.3-37 Revision 0

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.5 and SR 3.3.1.1.6 REQUIREMENTS (continued) These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status.

The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from the fully inserted position since indication is being transitioned from the SRMs to the IRMs.

The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the system design will prevent further increases (by initiating a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs and APRMs concurrently have onscale readings such that the transition between MODE 1 and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap between SRMs and IRMs similarly exists when, prior to withdrawing the SRMs from the fully inserted position, IRMs are above mid-scale on range 1 before SRMs have reached the upscale rod block.

As noted, SR 3.3.1.1.6 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap requirement has been met and indication has transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2).

(continued)

BFN-UNIT 1 B 3.3-38 Revision 0

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.5 and SR 3.3.1.1.6 (continued)

REQUIREMENTS If overlap for a group of channels is not demonstrated (e.g.,

IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel(s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition should be declared inoperable.

A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs.

SR 3.3.1.1.7 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 MWD/T average core exposure Frequency is based on operating experience with LPRM sensitivity changes.

SR 3.3.1.1.8 and SR 3.3.1.1.12 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.1.1.8 is based on the reliability analysis of Reference 9.

(continued)

BFN-UNIT 1 B 3.3-39 Revision 0

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.8 and SR 3.3.1.1.12 (continued)

REQUIREMENTS The 18 month Frequency of SR 3.3.1.1.12 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 shown that these components usually pass the Surveillance when performed at the 18 month Frequency.

SR 3.3.1.1.9, SR 3.3.1.1.10 and 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. For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, SR 3.3.1.1.13 includes physical inspection and actuation of the switches.

(continued)

BFN-UNIT 1 B 3.3-40 Revision 0

RPS Instrumentation RIPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.9, SR 3.3.1.1.10 and SR 3.3.1.1.13 (continued)

REQUIREMENTS Note 1 to SR 3.3.1.1.9 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 are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 MWD/T LPRM calibration against the TIPs (SR 3.3.1.1.7). A second Note for SR 3.3.1.1.9 is provided that requires the APRM and IRM SRs to be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of entering MODE 2 from MODE 1. Testing of the MODE 2 APRM and IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

The Frequency of SR 3.3.1.1.9 is based upon the assumption of a 92 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequency of SR 3.3.1.1.10 is based upon the assumption of a 184 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequency of SR 3.3.1.1.13 is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.

(continued)

BFN-UNIT I B 3.3-41 Revision 0

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 REQUIREMENTS (continued) The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function uses the recirculation loop drive flows to vary the trip setpoint. This SR ensures that the total loop drive flow signals from the flow units used to vary the setpoint are appropriately compared to a calibrated flow signal and, therefore, the APRM Function accurately reflects the required setpoint as a function of flow.

The Frequency of 18 months is based on system design considerations which do not support flow unit bypass during operation. Thus, this calibration is performed during refueling outages.

SR 3.3.1.1.14 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel.

The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function.

The 18 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 shown that these components usually pass the Surveillance when performed at the 18 month Frequency.

(continued)

BFN-UNIT 1 B 3.3-42 Revision 0

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 REQUIREMENTS (continued) This SR ensures that scrams initiated from the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is Ž 30% RTP. This involves calibration of the bypass channels (PIS-1-81A, PIS-1-811B, PIS-1-91A, and PIS-1-91B). Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.

If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at > 30% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are enabled), this SR is met and the channel is considered OPERABLE.

The Frequency of 18 months is based on engineering judgment and reliability of the components.

(continued)

BFN-UNIT 1 B 3.3-43 Revision 0

RPS Instrumentation B 3.3.1.1 BASES (continued)

REFERENCES 1. FSAR, Section 7.2.

2. FSAR, Chapter 14.

3. NEDO-23842, "Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.

4. FSAR, Appendix N.

5. FSAR, Section 14.6.2.

6. FSAR, Section 6.5.

7. FSAR, Section 14.5.

8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.

9. NEDC-30851-P-A,'Technical Specification Improvement Analyses for BWR Reactor Protection System,"

March 1988.

10. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

11. MED-32-0286, 'Technical Specification Improvement Analysis for Browns Ferry Nuclear Plant, Unit 2," October 1995.

BFN-UNIT I B 3.3-44 Revision 0

v t e Letter Sequence Request
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Initiation Request, Request Acceptance... Administration Questions 14 August 2006 Answers Results Approval Other: ML063180569
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ML062790140

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