ML042590554

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Supplemental Information Related to Revised Response to Generic Letter 94-02: Long-term Stability Solution
ML042590554
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 09/08/2004
From: Mckinney B
Susquehanna
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-94-002, PLA-5803
Download: ML042590554 (40)
v • d • e


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T.M-no Britt T. McKinney Vice President-Nuclear Site Operations SEP 0 8 2004 PPL Susquehanna, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3149 Fax 570.542.1504 btmckinney pplweb.com I

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P P 1444 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station OPI-17 Washington DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION SUPPLEMENTAL INFORMATION RELATED TO REVISED RESPONSE TO GENERIC LETTER 94-02:

LONG-TERM STABILITY SOLUTION PLA-5803 Docket Nos. 50-387 and 50-388

Reference:

1) PLA -5686, B. L SIriver (PPL) to USNRC, "Proposed Amendinent No. 259 to License NFP-14 and Proposed Amendment No. 224 to License NFP-22: Revised Response to CL 94-02: Long-Tenn Stability Solution, " dated December 22, 2003.

In PPL's December 22, 2003, proposed license amendment, (Reference 1), the core flow limit specified in the proposed SR 3.3.1.3.5 is 60 Mlb/Hr. During revision of the cycle specific Core Operating Limits Reports (COLR) for Susquehanna Units 1 and 2, it was determined that this core flow limit is non-conservative.

PPL has documented this discrepancy in the PPL Corrective Action process. The circumstances that led to this discrepancy will be evaluated and resolved in accordance with the requirements of that process.

Per discussion with the PPL NRC Project Manager, this supplement provides a revision to the information submitted in Reference I and is necessary for NRC to complete the review of Reference 1. The Enclosure to this letter provides the basis for the updated core flow limit. contains the markups of the U1 and U2 TS pages showing the revised core flow limit. contains the Camera Ready UI and U2 TS pages for the changes identified on the Attachment 1 markups. contains revised Ul and U2 TS Bases markups, provided for information only.

A001

2 Document Control Desk PLA-5803 A revised NSHC, which supersedes the NSHC in Reference 1, is included in Enclosure 1.

These changes do not involve a significant increase in the probability or consequences of an accident previously evaluated, do not create the possibility of a new or different kind of accident from any accident previously evaluated, and do not involve a significant reduction in a margin of safety.

There are no new commitments made in this letter. However, PPL is currently committed to implementing the OPRM trip function by September 30, 2004. If it is determined that the NRC approval of the license amendment will not occur as a result of this supplement by September 7, 2004, we request the final approval of the OPRM submittal include a condition stating that PPL shall implement the OPRM trip function within 90 days of approval of the change (instead of September 30, 2004). This will allow implementation of the of the OPRM trip function in accordance with the normal 13-week work management process.

If you have any questions, please contact Mr. Duane L. Filchner at (610) 774-7819.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on:

c B. T. McKinney

Enclosure:

PPL Evaluation of the Proposed Change Attachments: - Revised U1 and U2 TS Changes (Markups) - Revised U1 and U2 TS Changes (Camera Ready) - Revised Ul and U2 TS Bases Markups (Information Only) copy: NRC Region I Mr. A. J. Blarney, NRC Sr. Resident Inspector Mr. R. V. Guzman, NRC Project Manager Mr. R. Janati, DEP/BRP

Enclosure to PLA-5803 PPL Evaluation of the Proposed Change

Enclosure to PLA-5803 Page 1 of 5 TECHNICAL ANALYSIS This Technical Analysis is provided as supplemental information to the Reference 1 submittal.

The safety and efficacy of the installed system in meeting the design requirement of detecting and suppressing reactor core thermal-hydraulic instabilities is demonstrated and documented in several NRC reviewed and approved Licensing Topical Reports. Topical Report NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," Section 2.2, specifically states:

"The trip function will be enabled when both the power level is greater than 30%

of rated and the core flow is less than 60%. Based on experience with actual instabilities and decay ratio calculations, instabilities above 60% flow are not expected."

In correspondence between the BWROG and NRC dated September 7, 1996, the BWROG identified that the 60% core flow rate limit specified in the NEDO is to be interpreted as the core flow at 60% of original rated core flow. Thus the PPL core flow limit would have been specified as 60Mlb/Hr in the SR 3.3.1.3.5, since the original rated core flow for SSES Unit 1 and 2 was 100 Mlb/Hr. The BWROG followed up with a communication to all BWR's addressing the same interpretation.

However, on February 17, 2003, General Electric (GE) issued a communication to the BWROG Stability III Option plants that indicated the adequacy of the upper Armed Region (defined by the core flow criteria) should be evaluated before the OPRM is declared OPERABLE. This was deemed prudent because plant and cycle specific variations since the original report was issued could adversely affect stability performance such that the upper Armed Region may not provide the desired level of protection.

This evaluation was recently performed using the NRC approved methodology described in EMF-CC-074(P)(A), Volume 4. The methodology described in EMF-CC-074(P)(A),

Volume 4, has been used to establish the "immediate exit" region of the Susquehanna power flow maps, to assure that the plant is not operated in regions where instabilities could occur. As such, this reference is proposed to be retained in the Technical Specification Section 5.6.5 list of references for both Unit 1 and Unit 2.

Cycle specific calculations for both Unit 1 and 2 using this NRC approved methodology demonstrate that the OPRM should be enabled at 65 Mlb/hr for both U1C14 and U2C12 to ensure that the OPRM is enabled where an instability could occur. This flow rate is selected to correspond to the intersection of the highest rod line (APRM Rod Block Trip Setpoint) and the "Immediate Exit Region" (calculated by EMF-CC-074(P)(A)). The increase in the enabled region is needed because of higher energy loading due to use of twenty four month cycles and higher design capacity factors.

Enclosure to PLA-5803 Page 2 of 5 Based on this, the SR is proposed to be revised to read as follows:

"Verify OPRM is not bypassed when THERMAL POWER is 2 30% RTP and core flow < 65 Mlb/Hr."

Additionally, the NRC reviewed and approved methodology used to determine the upper Armed Region specified in the SR is to be retained in the Technical Specification Section 5.6.5 list of COLR references rather than removed as identified in Reference 1.

This reference is as follows:

lEMF-CC-074(P)(A), Vol. 4 l BWR Stability Analysis: Assessment of STAIF with

-- Input from MICROBURN-B2 With implementation of these proposed changes, the intent of the requirement of the NRC approved Topical Report NEDO-32465-A, Section 2.2, "Licensing Compliance" is met.

Revised No Significant Hazards Consideration:

The response to question 3 is revised (as identified by revision bar) to remove the reference to Topical Report NEDO-32465-A and to indicate that the protection from an instability event is demonstrated in accordance with NRC reviewed and approved Topical Reports without providing reference to specific Topical reports.

The revised No Significant Hazards Consideration Evaluation follows:

The Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The proposed amendment revises Susquehanna Steam Electric Station Unit 1 and Unit 2 Technical Specifications (TS). The change adds SSES Technical Specification (TS) 3.3.1.3, "Oscillation Power Range Monitor (OPRM)

Instrumentation," and revises TS 3.4.1 "Recirculation Loops Operating," and TS 5.6.5 "Core Operating Limits Report (COLR)."

Susquehanna, LLC (PPL) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three

Enclosure to PLA-5803 Page 3 of 5 standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1.

Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No.

The OPRM most directly affects the APRM and LPRM portions of the Power Range Neutron Monitoring system. Its installation does not affect the operation of these sub-systems. None of the accidents or equipment malfunctions affected by these sub-systems are affected by the presence or operation of the OPRM.

The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux changes. The APRM Fixed Neutron Flux-High function is capable of generating a trip signal to prevent fuel damage or excessive reactor pressure. For the ASME overpressurization protection analysis in FSAR Chapter 5, the APRM Fixed Neutron Flux-High function is assumed to terminate the main steam isolation valve closure event. The high flux trip, along with the safety/relief valves, limits the peak reactor pressure vessel pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis in Chapter 15 takes credit for the APRM Fixed Neutron Flux-High function to terminate the CRDA. The Recirculation Flow Controller Failure event (pump runup) is also terminated by the high neutron flux trip.

The APRM 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 Safety Limits (e.g., MCPR and Reactor pressure) being exceeded.

The installation of the OPRM equipment does not increase the consequences of a malfunction of equipment important to safety. The APRM and RPS systems are designed to fail in a tripped (fail safe) condition; the OPRM will have no affect on the consequence of the failure of either system. An inoperative trip signal is received by the RPS any time an APRM mode switch is 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. These functions are not specifically credited in the accident analysis, but are retained for the RPS as required by the NRC approved licensing basis.

Enclosure to PLA-5803 Page 4 of 5 The OPRM allows operation under operating conditions presently restricted by the current Technical Specifications by providing automatic suppression functions in the area of concern in the event an instability occurs. The consequences of any accident or equipment malfunction are not increased by operating under those conditions. Although protected by the OPRM from thermal-hydraulic core instabilities above 30% core power, operation under natural core circulation conditions is not allowed. No accidents or transients of a type not analyzed in the FSAR are created by operating under these conditions with the protection of the OPRM system.

This change does not increase the probability of an accident as previously evaluated. The OPRM is designed and installed to not degrade the existing APRM, LPRM, and RPS systems. These systems will still perform all of their intended functions. The new equipment is tested and installed to the same or more restrictive environmental and seismic envelopes as the existing systems. The new equipment has been designed and tested to electromagnetic interference (EMI) requirements, which assure correct operation of the existing equipment. The new system has been designed to single failure criteria and is electrically isolated from equipment of different electrical divisions and from non-lE equipment. The electrical loading is within the capability of the existing power sources and the heat loads are within the capability of existing cooling systems. The OPRM allows operation under operating conditions presently forbidden or restricted by the current Technical Specifications. No other transient or accident analysis assumes these operating restrictions.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different.

kind of accident from any accident previously evaluated?

Response: No.

This proposal does not create the possibility of a new or different type of accident from any accident previously evaluated. The OPRM system is a monitoring and accident mitigation system that cannot create the possibility for an accident not previously evaluated.

The OPRM will allow operation in conditions restricted by the current Technical Specifications. Although protected by the OPRM from thermal-hydraulic core instabilities above 30% core power, operation under natural circulation conditions is not allowed. No accidents or transients of a type not analyzed in the FSAR are created by operating under these conditions

Enclosure to PLA-5803 Page 5 of 5 with the protection of the OPRM system. No new failure modes of either the new OPRM equipment or of the existing APRM equipment have been introduced. Quality software design, testing, implementation and module self-health testing provide assurance that no new equipment malfunctions due to software errors are created. The possibility of an accident of a new or different type than any evaluated previously is not created.

The new OPRM equipment is designed and installed to the same system requirements as the existing APRM equipment and is designed and tested to have no impact on the existing functions of the APRM system.

Appropriate isolation is provided where new interconnections between redundant separation groups are formed. The OPRM modules have been designed and tested to assure that no new failure modes have been introduced.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

There has been no reduction in the margin of safety as defined in the basis for the Technical Specifications. The OPRM system does not negatively impact the existing APRM system. As a result, the margins in the Technical Specifications for the APRM system are not impacted by this addition.

Current operation under the ICAs provides an acceptable margin of safety in the event of an instability event as the result of preventive actions and Technical Specification controlled response by the control room operators.

The OPRM system provides an increase in the reliability of the protection of the margin of safety by providing automatic protection of the MCPR safety limit, while the protection burden is significantly reduced for the control room operators. This protection is demonstrated as described above, and in the referenced NRC reviewed and approved Topical Reports.

Replacement of the ICA operating restrictions from Technical Specifications with the OPRM system does not affect the margin of safety associated with any other system or fuel design parameter.

Therefore, this change does not involve a reduction in the margin of safety.

to PLA-5803 Revised Ul and U2 TS Changes (Markups)

PPL Rev. 0 OPRM Instrumentation 3.3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.3.5 Verify OPRM is not bypassed when THERMAL POWER 24 months is 2 30% RTP and core flow< P6MLb/Hr.

SR 3.3.1.3.6 NOTE Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within limits.

24 months on a STAGGERED TEST BASIS SUSQUEHANNA-UNIT 1 TS1/3.3-15c Amendment

PPL Rev. 0 OPRM Instrumentation.

3.3.1.3

.SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

-SR 3.3.1.3.5 Verfy OPRM is not bypassed when THERMAL POWER 24 months is 2 30% RTP and core flow.,WMLb1Hr.

SR 3.3.1.3.6 NOTE Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within limits.

24 months on a STAGGERED TEST BASIS SUSQUEHANNA - UNIT 2 TS / 3.3-1 5c

-Amendment

PPL Rev.2X Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

7. ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology l for Boiling Water Reactors EXEM BWVR Evaluation Model.'
8. XN-NF-79-71 (P)(A), "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors."
9. EMF-1997(P)(A), "ANFB-A0 Critical Power Correlation."
10. Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy l and Plant Safety While Increasing Operating Power Level Using the LEFM/7"m System," Engineering Report - BOP.
11. Caldon, Inc., "Supplement to Topical Report ER-80P: 'Basis for a Power Uprate with the LEFMI/'1m or LEFM CheckPlus"' System,",

Engineering Report ER -160P.

12. EMF-85-74(P)(A), "RODEX 2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model."
13. EMF-CC-074 (P)(A), Volume 4, "BWR Stability Analysis:. Assessment l

of STAIF with Input from MICROBURN-B2."*

14. EMF-2158(P)(A), OSiemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2."
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Lj F5po z# o 5. A*

W tZ 4 I&A d eier A

re SAL Isoeere-ss So/D&he LLCe+,jj 8X1u Mehdology 4i~- t~d A d I (continued)

SUSQUEHANNA - UNIT 1 TS / 5.0-23 A end ent 8 10i, 1ix9, A4,,209, X5 z

P'9L AI.

Reporting Requirements

- 5.6 5.6 Reporting Requirements 5.6.5 QOLR (continued)

11.

Caldon, Inc., 'Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMv'/ or LEFM CheckPlusm System," Engineering Report ER-160P.

12.

EMF-85-74(P)(A), RODEX 2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model."

13:

EMF-2158(P)(A), 'Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO4lMicrobum-B2,"

Siemens Power Corporation.

C.

14.

EMF-CC-074(P)(A), Volume 4, BWR Stability Analysis: Assessmet 'of STAIF with Input from MICROBURN-B2.f The core operating limits shall be determined such that all applicable limits (e.g.,

fuel thermal mechanical limits, core thermal hydrauliclimits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met I

5.6.6

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reiprtffwithin 30 days.

Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revisibn 3, Regulatory Position C.4.

PAM Report When a report is required by Condition B or F of LCO 3.3.3.1, 'Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 IS. W EDo 3Z4oS--A, /

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¶,ffg-ess So hxir^M LtcJ.-s G Meait o46kogy NA - UNIT 2 TS 1 5.0-23 Amendment 1l SUSQUEHAN 1,P4, I

to PLA-5803 Revised Ul and U2 TS Changes (Camera Ready)

PPL Rev. 0 OPRM Instrumentation 3.3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.3.5 Verify OPRM is not bypassed when THERMAL POWER 24 months is 2 30% RTP and core flow < 65 MLbIHr.

SR 3.3.1.3.6 NOTE Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within limits.

24 months on a STAGGERED TEST BASIS I

SUSQUEHANNA - UNIT 1 TS / 3.3-15c Amendment

PPL Rev. 0 OPRM Instrumentation 3.3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.3.5 Verify OPRM is not bypassed when THERMAL POWER 24 months is 2 30% RTP and core flow S 65 MLb/Hr.

SR 3.3.1.3.6 NOTE -------

Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within limits.

24 months on a STAGGERED TEST BASIS I

SUSQUEHANNA - UNIT 2 TS / 3.3-1 5c Amendment

PPL Rev.

Reporting Requirements 5.6.

5.6 Reporting Requirements 5.6.5 COLR (continued)

.7. ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model."

8. XN-NF-79-71 (P)(A), Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors.
9. EMF-1 997(P)(A), 'ANFB-10 Critical Power Correlation.
10. Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFMO/TM System," Engineering Report - 80P.
11. Caldon, Inc., 'Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM/TM or LEFM CheckPlusTu System,", Engineering Report ER -160P.
12. EMF-85-74(P)(A), "RODEX 2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model."
13. EMF-CC-074 (P)(A), Volume 4, "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2."
14. EMF-21 58(P)(A), 'Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2."
15. NEDO-32465-A, -BWROG Reactor Core Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications."
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

SUSQUEHANNA - UNIT 1 TS / 5.0-23 Amendment 8 1 V6, 1, 9. fi 4,,l09, V6 '6

PPL Rev.

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

11.

Caldon, Inc., 'Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMvM or LEFM CheckPiusTm System," Engineering Report ER-160P.

12.

EMF-85-74(P)(A), "RODEX 2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model."

13.

EMF-2158(P)(A), 'Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/Microbum-B2,"

Siemens Power Corporation.

14.

EMF-CC-074(P)(A), Volume 4, "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN B2."

15.

NEDO-32465-A, "BWROG Reactor Core Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications."'

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.

Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4.

(continued)

SUSQUEHANNA - UNIT 2 TS / 5.0-23 Apnoment 1 V/9 1 PAt6, 1 M.

PPL Rev.

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 PAM Report When a report is required by Condition B or F of LCO 3.3.3.1, 'Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

SUSQUEHANNA - UNIT 2 TS / 5.0-23a Ampnoment 49 1 P 4, 196 to PLA-5803 Revised U1 and U2 TS Bases Markups (Information Only)

.,r Information Only OPRM Instrumentatlon 8 3.3.1.3 8 3.3 INSTRUMENTATION 8.3:3.1.3 Oscillation Power Range Monitor (OPRM)

BASES BACKGROUND General Design Criterion 10 (GDC 10) requires the reactor core and associated coolant control and protection systems to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition of normal operation Including the affects of anticipated operational occurrences.

Additionally, GDC 12requires the reactor core and associated coolant control and protection systems to be designed to assure that power.

oscillations which can result in conditions exceeding acceptable fuel design.limits are either not possible or can be reliably and readily detected and suppressed' The OPRM System Provides compliance with GDC 10 and GDC 12 thereby providing protection from exceeding the fuel MCPR safety limit.

References 1, 2. and 3 describe three separate algorithms for detecting stability related oscillations: the period based detection algorithm, the amplitude based algorithm. and the growth rate algorithm. The OPRM System hardware implements these algorithms in microprocessor based modules. These modules execute the algorithms based on LPRM inputs and generate alarms and trips based on these calculations. These trips result In tripping the Reactor Protection System (RPS) when the appropriate RPS trip logic is satisfied, as described In the Bases for LCO 3.3.1.1. RPS Instrumentatioi. Only the period baseddetetion algorithm is used in the safety analysis (Ref.1, 2 6, & 7). The remaining algorithms provide defense-in-depth and additional protection against unanticipated oscillatons.

(continued)

-r3/6S.; k SUSQUEHANNA UNIT 1 12je0 vii

  • O

'or Information Only a,

OPRM Instrumentation B 3.3.1.3 BASES BACKGROUND The period based detection algorithm detects a stability-(continued) related oscillation based on the occurrence of a fixed number of consecutive LPRM signal period confirmations followed by the LPRM sigpoal amplitude exceeding a specified setpoint. Upon detection of a stability related oscillation a trip is generated for that OPRM channel.

The OPRM System consists of 4 OPRM trip channels, each channel consisting of two OPRM modules. Each OPRM module receives input from LPRMs. Each OPRM module also receives, Input from the NMS average power range monitor (APRM) power and flow signals to automatically enable the trip function of the OPRM module.

Each OPRM module is continuously tested by a self-test function. On detection of any OPRM module failure, either a Trouble alarm or INOP alarm Is activcted. The OPRM module provides an INOP alarm when the self-test feature Indicates that the OPRM module may not be capable of meeting its functional requirements APPLICABLE It has been shown that BWR cores may exhibit thermal-hydraulic SAFETY ANALYSES reactor instabilities in high power and low flow portions of the core power to flow operating domain. GDC 10 requires the reactor core and associated coolant control and protection systems to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition of normal operation, Including fe

<,offects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding acceptable fuel design limits are either not possible or can be

.reliably and readily detected and suppressed. The OPRM System.

provides compliance with GDC 10 and GDC 12 by detecting the onset of oscillations and suppressing them by initiating a reactor scram. This assures that the MCPR safety limit will not be violated for anticipated oscillations.

The OPRM Instrumentatidn satisfies Criterion a of the NRC Policy Statement Ts N

SUSQUEHANNA UNIT I

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-For Information Only i rL I-V.

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OPRM Instrumentation B 3.3.1.3 BASES LCO

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T518i Four channels of the OPRM System are required to be OPERABLE to ensure that stability related oscillations are detected and suppressed prior to exceeding the-MCPR safety limit Only one of the two OPRM modules' pcrouo 4 b3fd eteest.

trs required for OPRM channel OPERABILITY. The minimum number of LPRMs required OPERABLE to maintain an OPRM channel OPERABLE Is consistent with the minimum number of LPRMs required to maintain the APRM system PPERABLE per LCO 3.3.1.1.

~ V ~ J ~

p f h e ' Q R M ~ r i o.~ p e d APPLIC. AILITY The OPRM instrumentation is required to be OPERABLE in order to detect and suppress neutron flux oscillations in the event of thermal.-

hydraulic instability. As described In References 1, 2, and 3, the power/core flow region protected against anticipated oscillationsaIJ defined by THERMAL POWER 2 30% RTP and core flow :5AMlb/Hr.

The OPRM trip is required to be enabled In this region, and the OPRM must be capable of enabling the trip function as a result of anticipated transients that place the core in that power/flow condition. Therefore, the OPRM is required to be OPERABLE with THERMAL POWER 2 25%

RTP and at all core flows while above that THERMAL POWER. It is not necessary for the OPRM to be operable with THERMAL POWER < 25%

RTP because transients from below this THERMAL POWER are not anticipated to result In power that exceeds 30% RTP.

ACTIONS A Note has been provided to modify the ACTIONS related to the OPRM Instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered subsequent divisions, subsystems, components, or variables expressed in the Conditioti 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 OPRM 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 OPRM instrumentation channel.

(continued)

. -175 /:B 1.3 - 4 sc-13 8.8_1XX_

SUSOUEHANNA UNIT I 4VIII, o

Insert TSB1 The OPRM setpoints are determined based on the NRC approved methodology described in NEDO-32465-A (Ref. 6). The Allowable Value for the OPRM Period Based Algorithm setpoint (SP) is derived from the analytic limit corrected for instrument and calibration errors as contained in the COLR.

The OPRM bypass flow setpoint (SR 3.3.1.3.5) is conservatively established based on the greater of 60 Mlb/Hr. (NEDO-32465-A) and the value obtained based on the NRC approved methodology described in EMF-CC-074(P)(A),-Volume 4, (Reference 11).

Nor Information only I r

  • OPRM Instruneniatlon B 3.3.1.3 ACTIONS (continued)

A 1,A 2. andA3 Because of the reliability and on-line self-testing of the OPRM instrumentation and the redundancy of the RPS design, an allowable out of service time of 30 days has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status.

However, this out of service lime is only acceptable provided the OPRM instrumentation sill maintains OPRM trip capability (refer to Required Actions 8.1 and 1.2). The remaining OPERABLE OPRM channels continue to provide trip capability (see Condition 6) and provide operator information relative to stability activity. The remaining OPRM modules have high reliability. With this high reliability, there Is a low probability of a subsequent channel failure within the allowable out of service time. In" addition, the OiRM modules continte to perform on-line self-testing end alert the operator if any further system degradation occurs.

It the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the OPRM channel or associated RPS trip system must be placed in the trpped condition per Required Actions A.1 and A2. Placing the inoperable OPRM channel in trip (or the associated RPS trip system in trip) would conservatively compensate for the inopeiability, provide the capability to accommodate a single failure and allow operation tb continue. Altemately, If It is not desired to place the OPRM channel (or RPS trip system) in trip (e.g.. as in the case where placing the inoperable channel in trip would result In a full scram).

the altemale method of detecting and suppressing thermal hydraulic instability oscillations is required (Required Action A.3). This alternate method is described in Reference 5. It consists of increased operator awareness and monitoring for neutron flux oscillations when operating In the region where oscillations are possible. If indications of oscillation, as described In Reference 5 are observed by the operator, the operator will take the actions described by procedures which Include Initiating a manual scram of the reactor. fli e-foie /.Plow r),utj A

1j4OSOIC\\a%

ga.RJelrj As 4ar e l\\.

,e c0y11CALL.C Y4?L (l< C~intl 1 he COLA.

-T3 46 3 - 3d.

SUSOUEHANNA UNIT 1

  • I.Qwg o 0

.or Information uniy OPRM Instrumentrtion B 3.3.1.3 ACTIONS B.1 and 8.2 (continued)

Required action 8.1 Is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped OPRM channels within the same RPS trip system result in not maintaining OPRM trip capability. OPRM trip capability Is considered to be maintained when sufficient OPRM channels are OPERABLE or in trip (or the associated RPS trip system is in trip), such that a valid OPRM signal will generate a trip signal In both RPS trip systems (this would require both RPS trip systems to have at least one OPRM channel OPERABLE or the associated RIPS trip system In trip).

Because of the low probability of the occurrence of an instability, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Is an acceptable time to initiate the alternate method of detecting and suppressing thermal hydraulic instability oscillations described in Action A.3 abo-ve. The altomate method of detecting and suppressing thermal hydraulic instability oscillations would adequately address detection and mitigation in the event of instability oscillations. Eased on industry operating experience with actual instability oscillations, the operator would be able to recognize instabilities during this time and take action to suppress them through a manual scram. In addition, the OPRM System may still be available to provide alarms to.the operator if the onset of oscillations were to occur. Since plant operation Is minimized in areas where oscillations may occur, operation for.120 days without OPRM trip capability is considered acceptable with implementation of the alternate method of detecting and suppressing thermal hydraulic instability oscillations, C.1 With any Required Action and associated Completion Tine not met, THERMAL POWER miust be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Reducing THERMAL POWER to < 25% RTP places the plant in a condition where instabilities are not likely to occur. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience, to reduce THERMAL POWER c 25% RTP from full power conditions in an orderly manner and without challenging plant systems.

-5

-UN U I3.3

- 43e SUSQUEHANNA UNIT 1Ie

'r Information Only OPRM Instrumentatlon B 3.3.1.3 SURVEILLANCE SR 3.3.1.3.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed to ensure that the entire channel will perform the intended function. A Frequency of 184 days provides an acceptable level of system average availability over the Frequency and is based on the.refiability of the channel (Ref. 7).

SR 3.3.1.3.2 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 OPRM System. The 1000 MWDIMT Frequency is based on overating experience with LPRM sensitivity changes.

SR 3.3.1.3.3 A CHANNEL CALIBRATION 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. Calibration of the channel provides a check of the internal reference voltage and the internal processor clock frequency. It also compares the desired trip setpoints with those in processor memory. Since the OPRM Is a digital system, the internal reference voltage and processor clock frequency are, In turn, used to automatically calibrate the Internal analog to digital converters. As noted, neutron detectors are exduded 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 1000 MWD1MT LPRM calibration using the TIPs (SR 3.3.1.3.2).

'SUSQUEHANNA UNIT 1

or Information unly OPRM Instrumentation 8 3.3.1.3 SURVEILLANCE REQUIREMENTS The Frequency of 24 months is based upon the assumption of (continued) the magnitude of equipment drift provided by the equipment supplier. (Ref. 7)

SR 3.3.1.3.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the

-OPERABILITY of the required trip logic for a specific channel; The functional testing of control rods, in LCO 3.1.3, Control Rod OPERABILITY, and scram discharge volume (SDV) vent and drain valves, in LCO 3.1.8, bScram Discharge Volume (SDV) Vent and Drain Valves', overlaps this Surveillance to provide complete testing of the assumed safety function. The OPRM self-test function may be utilized to perform this testing for those components that al is design.d to monitor.

The 24 month Frequency is based on engineering judgment, reliability of the components and operating experience.

SR 3.3.1.3.5 This SR ensures tt9 trips initiated'from the OPRM System will not be inadvertently bypssed when THERMAL POWER Is 2 30% RTP and core flow iss WMLbIHr. This no~rmally involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodology are incorporated into the actual setpoints (Reference 7).

If any bypass chanriel setpoint is nonconservative (.e., peOPRM module is bypassed at Ž 30% RTP and core flow Is StLbMlr), then the affected OPRM module is considered inoperable. Alternatively, te bypassed channel can be manually placed In the conservative position (Manual Enable). If placed in the MANUAL ENABLE condition. this SR Is met and the module is considered OPERABLE.

The 24 month Frequency Is based on engineering judgment and reliability of the components.

SUSQUEHANNA UNIT 1

  • E_4-X

,or Information uniy OPRM Instrumentdllon 6 3.3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.3.6 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the safety analysis (Ref. 6).

The OPRM self-test function may be utilized to perform this testing for those components it is designed to mnonitor. The'LPRM amplifier cards:

inputting to the OPRM are excluded from the OPRM RESPONSE TIME testing. The RPS RESPONSE TIME acceptance criteria are Included in Reference S.

As noted, neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. RPS RESPONSE TIME tests are conducted on an 24 mionih STAGGERED TEST BASIS. This Frequency is based upon operating experience, which shows that random failures of Instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.

SUSQUEHANNA UNIT I Ts/a3 3?,-431%

44-IsS '

'or Information oniy OPRM Instrumentation B 3.3.

1.3 REFERENCES

1. NEDO-3196MA, 'BWR Owners Group Long-Term Stabirity Solutions Licensing Methodology", November 1995.
2. NEDO 31960-A, Supplement 1 "3WR Owners Group Long-Term Stability Solutions Ucensing Methodology", November 1995.
3. NRC Letter, A. Thadani to L A. England, 6Acceptence for.

Referencing of.Topical Reports NEDO-31960, Supplement 1. '8WR Owners Group Long-Termn Stability Solutions Licensing Methodology, July 12, 1994.

4. Generic Letter 902, "Long-Teim Solutions and Upgrade of Inteiim Operating Recommendations for Thermal-Hydraulic Instabilries in
  • elli!ng Watcr Reactors", July 11. 1994.
5. BWROG Letter BWROG-9479, Guidelines for Stability Interim Corrective Action", June 6. 1994.
6. NEDO-32465-A, 'BWR Owners Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications", August 1996.
7. CENPD-400-P-A, Rev 01, "Generic Topical Report for the ABB Option Ill Oscillation Power Range Monitor (OPRMr, May 1995.
  • 8. FSAR, Table 7.3-28.

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Information Only 1-,

_1 A# h OPRM Instrumentation

' B 3.3.1.3 633.3 INSTRUMENTATION

.3.3.1.3 Oscillation Power Range Monitor (OPRM)

BASES BACKGROUND General Design Criterion 10 (GDC 10) requires the reactor core and associated coolant control and protection systems to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any conditon of normal operation

Additionally, GDC 12 -requires the reactor core and associated coolant control and protection systems to be designed to assure that power osciflations which can result in conditions exceeding acceptable fuel design.limits are either not possible or can be reliably and readily detected and suppressed. The OPRM System provide* corpo.ance with GDC 10 and GDC 12 thereby providing protection from exceeding the

' fuel MCPR safety limit.

References 1, 2, and 3 describe three separate algorithms for detecting stability related oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. The OPRM System hardware implements these algorithms in microprocessor based modules. These modules execute the algorithms based on LPRM inputs and generate alarms and trips based on these calculations. These trips result In tripping the Reactor Protection System (RPS) when te appropriate RPS trip logic Is satisfied, as described In the Bases for LCO

  • 3.3.1.1. 'RPS Instrumentatiohf Only the period based detection algorithm is used in the safety analysis (Ref. 1, 2, 6, & 7). The remaining algorithms provide defense-In-depth and additional protection against unanticipated oscillations.

(continued)

-r /64, k SUSQUJEHANNA UNITA* 2-W131-0

or Lnroxu2ILL.I-u o

_,_j OPRM Instrumentation

.6 3.3.1.3 BASES BACKGROUND The period based detection algorithm detects a stability-(continued) related oscilation based on the occurrence of a fixed number of consecutive LPRM signal period confirmations followed by the LPRM signal amplitude exceeding a specified sitpoint Upon detection of a stability related oscillation a top is generated for that OPRM channel.

The OPRM System consists of 4 OPRM trip channels, each channel consisting of two OPRM modules. Each OPRM module receives input from LPRMs. Each OPRM module also receives Input from the NMS average power range monitor (APRM) power and flow signals to automatically enable the trip function of the OPRM module..

Each OPRM module is continuously tested by a self-test function. On detection of any OPRM module failure, either a Trouble alarm or INOP alarm Is zctivzted. The OPRM module provides an INOP alarm when the self-test feature indicates that the OPRM module may not be capable of meeting its functional'requirements APPLICABLE It has been shown that BWR cores may exhibit thermal-hydraulic SAFETY ANALYSES reactor instabilities in high power and low flow portions of the core power to flow operating domain. GDC 10 requires the reactor core and associated coolant control and protection systems to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition of normal operation, includin the erffetS of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. The OPRM System provides compliance with GtDC 10 and GDC 12 by detecting the onset of oscillations and suppressing them by initiating a reactor scram. This assures that the MCPR safety limi will not be violated for anticipated oscillations.

The OPRM Instrumentaticin satisfies Criterion 3 of the NRC Policy Statement SUSQUEHANNA UNITA 7-

' -,nformation Only OPRM Instrumentation B 3.3.1.3 BASES LCO c-.

Four channels of the OPRM System are required to be OPERABLE to ensure that stability related oscillations are detected and suppressed prior to exceeding the-MCPR safety limit Only one of the two OPRM modules'pered4 d

t is required for OPRIMI channel OPERABILIMY. The minimum number of LPRMs required OPERABLE to maintain an OPRM channel OPERABLE Is consistent with the minimum number of LPRMs required to maintain the APRM system OPERABLEi Opr LCO 3.3.1.1.

we,;Sll le V 8-for OP P rio sedl Aoti mpon is rid e

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,APPLICABILITY-The OPRM Instrumentation is required to be OPERABLE in order to detact and suppress neutron flux oscillations in the event of thermal-.

hydraulic instability. As described In References 1, 2, and 3, the

)

power/core flow region protected against anticipated osclle l

defined by THERMAL POWER 2 30% RTP and core flow : 5(Mlb/Hr.

The OPRM trip is required to be enabled In this region, and the OPRM must be capable of enabling the tip function as a result of anticipated transients that place the core in that powerltlow condition. Therefore, the OPRM Is required to be OPERABLE with THERMAL POWER 2 25%

RTP and at all core flows while above that THERMAL POWER. It Is -not necessary for the OPRM to be operable with THERMAL POWER < 25%

RTP because transients from below this THERMAL POWER are not anticipated to result In power that exceeds 30% RTP.

ACTIONS A Note has been provided to modify the ACTIONS related to the OPRM 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 wilt 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 OPRM 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 OPRM Instrumentation channel.

(continued) vri/asa- 0c

.13 8.8 XX SUSQUEHANNA UNIT4 Z

Insert TSB1 The OPRM setpoints are determined based on the NRC approved methodology described in NEDO-32465-A (Ref. 6). The Allowable Value for the OPRM Period Based Algorithm setpoint (SP) is derived from the analytic limit corrected for instrument and calibration errors as contained in the COLR.

The OPRM bypass flow setpoint (SR 3.3.1.3.5) is conservatively established based on the greater of 60 Mlb1Hr. (NEDO-32465-A) and the value obtained based on the NRC approved methodology described in EMF-CC-074(P)(A), Volume 4, (Reference I1).

or Information. Only OPRM InsrNmenrtalon B 3.3.1.3 ACTIONS (continued)

A 1.A 2. and A 3 Because of the reliability and on-line self-testing of the OPRM Instrumentation and the redundancy of the RPS design, an allowable out of service time of 30 days has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status.

However, this out of service time is only acceptable provided the OPRM

-instrumentatlon-still maintains OPRM trip capableity (refer to Required Actions 1.1 and B.2). The remaining OPERABLE OPRM channels continue to provide trip capability (see Condition B) and provide operator information relative to stability activity. The remaining OPRM modules have high reliability. With this high reliability, there Is a low probability of a subsequent channel failure within the allowable out of service time. In addition, the OPRM modules continue to -perforrn onirns self-tPiting end alert the operator if any further system degradation o

wrs.

If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the OPRM channel or associated RPS trip system must be placed in the tripped condition per Required Actions A.: and A.2. Placing the inoperable OPRM channel in trip (or the associated RPS trip system in trip) would conservatively compensate for the inoperability, provide the capability to accommodate a single failure and allow operation tb continue. Alternately, if It is not desired to place the OPRM channel (or RPS trip system) in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram).

the alternate method of detecting and suppressing thermal hydraulic instability oscillations is required (Required Action A.3). This alternate method is described in Reference 5. It consists of increased operator awareness and monitoring for neutron flux oscillations when operatng In the region where oscillations are possible. If indications of oscillation, as described In Reference 5 are observed by the operator, the operatorwil take the actions described by procedures which Include Initiating a manual scram of the reactor. -Me fe Y /4h'.4

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rInformation Only OPIRM Instrumentation

.8B 3.3.1.3 ACTIONS B.1 and 8.2

-(continued)

Required action B.1 Is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped OPRM channels within the same RPS trip system result in not maintaining OPRM trip capability. OPRM trip capability is considered totbe maintained when sufficient OPRIM channels are OPERABLE or in trip (or the associated RPS trip system is in trip), such that a valid OPRM signal will generate atrip signal in both RPS trip systems (this would require both RPS trip systems to have at least one OPRM channel OPERABLE or the associated RPS trip system In trip).

Because of the low probability of the occurrence of an instability, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is an acceptable time to initiate the alternate method of detecting and suppressing thermal hydraulic instability oscillations described In Action A.3 above. The altdriurxe tnethod of detecting and suppressing' thermal hydraulic instability osciliations would adequately address detection and mitigation in the event of instability oscillations. Based on industry operating experience with actual instability oscillations, the operator would be able to recognize instabilities during this time end take action to suppress them through a manual scram. In addition, the OPRM System may still be available to provide alarms to the operator it the onset of oscillations were to occur. Since plant operation is minimized in areas where osciliations may occur, operation for 120 days without OPRM trip capability is considered acceptable with implementation of the alternate method of detecting and suppressing thermal hydraulic instability oscillations.

With any Required Action and associated Completion Time not met; THERMAL POWER must be reduced to < 25% RTP witin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Reducing THERMAL POWER to c 25% RTP places the plant In a condition where instablNies are not likely to occur. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience, to reduce THERMAL POWER < 25% RTP from full power conditions in an orderly manner and without challenging plant systems.

SUSQUEHANNA UNIT.* Z-

  • Anformation Only A

roL sa an OPRM Instrumentation B Z.3.1.3 SURVEILLANCE REQUIREMENTS SR 3.3.1.3.1 A CHANNEL FUNCTIONAL TEST is performed to ensure that the entire channel will perform the intended function. A Frequency of 184 days provides an acceptable level of system average availability over fte Frequency and Is based on the-reliability of the channel (Ref. 7).

SR 3.3.1.3.2 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 inkut to the OPRM System. The 1000 MWD/MT Frequency is based on operating-ezporisrct-wftr LPRM. sonsitivity changes.

SR 3.3.1.3.3 A CHANNEL CALIBRATION 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. Calibration of the channel provides a check of the internal reference voltage and ht internal processor cl6ck frequency. It also compares the desired trip setpoints with those in processor-memory. Since the OPRM Is a digital system, the internal reference voltage and processor clock frequency are, in turn, used to automatically calibrate the internal analog to digital.

converters. As noted, neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal driflt and because of the difficulty of simulating a meaningful signal. Changes In neutron detector sensitivity are compensated for by performing the 1000 MWD/MT LPRM calibration using the TIPs (SR 3.3.1.32).

SUSQUEHANNA UNIT

  • 2 T3/83---414

. B 9.94(*

S.

a.

OPRM Instrumentation B 3.3.1.3 SURVEILLANCE REQUIREMENTS The Frequency of 24 months is based-upon the assumption of (continued) the magnitude of equipment drift provided by the equipment supplier. (Ref. 7)

SR 3.3.1.3.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods, In LCO 3.1.3, 'Control Rod OPERABILITY", and scram discharge volume (SDV) vent and drain valves, in LCO 3.1.8, 'Scram Discharge Volume (SDV) Vent and Drain Valves', overlaps this Surveillance to provide complete testing of the assumed safety function. The OPRM self-test function may be utmized to:

perform this testing for those components.thet it is designed to monitor.

The 24 month Frequency is based on engineering judgment, reliability of.

the components and operating experience.

SR 3.3.1.3.6 This SR ensures trips initiated-from the OPRM System wll not be inadvertently bypssed when THERMAL POWER is 2 30% RTP and core flow is s LMLbIHr. This normally Involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodology are incorporated into the actual setpoints (Reference 7).

If any bypass channel setpoint is nonconservative (i.e.,

module Is bypassed at 2 30% RTP and core flowis s sMLbM*. then the affected OPRM module Is considered inoperable. Altea ty, te bypassed channel can be manually placed in the conservative position (Manual Enable). If placed In the MANUAL ENABLE condition, tis SR s met and the module is considered OPERABLE.

The 24 month Frequency Is based on engineering judgment and reliability of the components.

SUSOUEHANNA UNIT> Z-

-- 3,X

.1

c Information Ozily

  • I OPRM Instrurventelson B 3.3.1.

SURVEILLANCE REOUIREMENTS (continued SR 3.3.1.3.6 This SR ensures that the Individual channel response times are less than or equal to the maximum values assumed In the safety analysis (Ref. 6).

The OPRM self-test function may be utilized to perform this testing for those components It is designed to monitor. The LPRM amplifier cards Inputting to the OPRM are excluded from the OPRM RESPONSE TIME testing. The RPS RESPONSE TIME acceptance criteria are Included in Reference 8.

As noted, neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. RPS RESPONSE TIME tests are, conducted on an 24 month STAGC-ERED TEST BASIS. Thas Frequency is based upon operating experience, which shows that random failures of Instrumentation components causing serious response time degradation, but not channel failure, are Infrequent occurrences.

m-/'3 3.:-4A Lh

-SUSQUEHANNA UNIT*.-i.

IZ saS.

jgnformation Only 4

rrL-7W-r. c OPRM Instrumentation B 3.3.

1.3 REFERENCES

1. NEDOw31960-A, -BWR Owners Group Long-Term Stability Solutions Licensing Methodology",.November 1995.
2. NEDO 31960-A, Supplement 1 "BWR Owners Group Long-Term Stability Solutions Ucensing Methodologyt, November 1995.

.t

3. NRC Letter, A. Thadani to L A. England, 'Acceptance for Referencing of Topical Reports NEDO-31960. Supplement 1, 'BWR Owners Group Long-Term Stability Solutions Licensing Methodology.

July 12, 1994.

4. Generic Letter 94-02, "Long-Teirn Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities In Boiling Wdaor Reactlar", July 11. 1994.
5. BWROG Letter BWROG-9479, *Guidelines for Stabifity Interim Corrective Action', June 6. 1994.
6. NEDO-32465-A, -BWR Owners Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications", August, 1996.
7. CENPD-400-P-A, Rev 01. 'Generic Topical Report for the ABB Option IlIl Oscillation Power Range Monitor (OPRM)'. May 1995.
8. PSAR, Table 7.3-28.
10.

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