ML18033A805

Proposed Tech Specs,Adding Section 3.5/4.5-M & Revising Section 3.6.f to Provide Limiting Conditions for Operation & Surveillance Requirements for Reactor Core thermal-hydraulic Stability

ML18033A805
Person / Time
Site:	Browns Ferry
Issue date:	06/20/1989
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18033A804	List: ML18033A803 ML18033A805
References
NUDOCS 8906280080
Download: ML18033A805 (26)

Text

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATIONS REVISIONS BROWNS FERRY NUCLEAR PLANT UNIT 2 (TVA BFN TS 272) 8906280080 o 5000260 PDR PgOCK pDG P

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Section a eÃo.

D. Reactivity Anomalies 3.3/4.3-11 E. Reactivity Control 3.3/4.3-3.2 F. Scram Discharge Volume 3.3/4.3-12 3.4/4.4 Standby Liquid Control System. 3.4/4.4-1 A. Normal System Availability . 3.4/4.4-1 B. Operation with Inoperable Components 3.4/4.4-2 C. Sodium Pentaborate Solution. 3.4/4.4-3

3. 5/4. 5 Core and Containment Cooling Systems 3.5/4.5-1 A. Core Spray System (CSS). 3.5/4.5-1 B. Residual Heat. Removal System (RHRS)

(LPCZ and Containment Cooling) 3.5/4.5-4 C. RHR Service Water System and Emergency Equipment Cooling Water System (EECWS} 3.5/4.5-9 D. Equipment Area Coolers 3.5/4.5-13 E. High Pressure Coolant Injection System (HPCIS). 3. 5/4. 5-13 F. Reactor Core Isolation Cooling System (RCICS). 3.5/4.5-14 G. Automatic Depressurization System (ADS). 3.5/4.5-15 H. Maintenance of Filled Discharge Pipe . 3.5/4.5-17 I. Average Planar Linear Heat Generation Rate . . 3.5/4.5-18 J. Linear Heat 'Generation Rate'LHGR) 3.5/4.5-18 K. Minimum Critical Power Ratio (MCPR). 3.5/4.5-19

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L. APRM Setpoints . 3.5/4.5-20 M. Core Thermal-Hydraulic Stability 3.5/4.5-20 3.6/4.6 Primary System Boundary. 3..6/4.6-1 A. Thermal and Pressurization Limitations 3.6/4.6-1 B. Coolant Chemistry. 3.6/4.6-5 BFN Unit 2 ii

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2.1.1 APRM Flov Reference Scram and APRM Rod Block Sett'ngs . . . . . . . . . . . . . . . . . . . . 1.1/2.1-6 2.1-2 APRM Flov Bias Scram Vs. Reactor Core Flov... '. l.l/2.1-7 4.1-1 Graphic Aid in the Se&ction of an Adequate Interval Between Tests 3.1/4.1-13 4.2-1 System Unavai'ability. 3.2/4.2-64 3.5.K-1 MCPR Llmlts ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ 3.5/4.5-22 3.5.M-1 BFiV Power/Flow Stability Regions 3.5/4.5-22a 3.5.2 K f Factor. 3.5/4.5-23 3.6-1 Minimum Temperature 'F Above Change in Transient Temperature. 3.6/4.6-24 3.6-2 Change in Charpy V Transition Temperature Vs.

Neutron Exposure . 3.6/4."6-25 4.8.1.a Gaseous Release Points and Elevations 3.8/4.8-10 4.8.1.b Land Site Boundary . 3.8/4.8-11 6.2-1 Offsite Organization for Facility Management and Technical Support ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-33 6.2-2 Facility Organization 6.0-34

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/4.< COR~ AND CO. ~I<FK'iT CCOLiifG SVSTEMS LIMITIHG COHDITIONS F R OPERATION SURVEILLANCE REQUI REiiEiiTS 3.5 Core and Containment Coo i Svst ms 4.5 Core and Containment Cool'..o Systems L. APRM Setto'nts

1. whenever the core thermal FRP/CiiLPD shall be power is 2. 25/ of rated, the determined daily vhen ration of FRP/CNFLPD shall the reactor is 2. 25/ of be 2, 1.0, or the APRM scram rated thermal power.

and rod block setpoint equations listed in Sections 2.1..A and 2.1.B shall be multiplied by FRP/CPZLPD as follows:

Sq (0.66W + 54=) FRP QiLPD

~ (0 66M + 427) (gRP CiiLPD

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2. Shen it is determined that 3.5.L.l is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is alloved to correct the condition.

3. If 3.5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to g 25/ of rated thermal pover vithin 4 hours.

M. Core e~al-H draulic" Stabil t M. Core Thermal-H draulic Stabilit

1. The reactor shall not be 1. Verify that the reactor is operated at a thermal power outside of Region I and II and core flow inside of of Figure 3.5;M-1:

Regions I and II"of Figure 3.5.M-l. a. Following any increase of more than 5/ rated

2. If Region I of Figure 3.5.M-1.. thermal power vhile initial is entered, immediately core flow is less than initiate a manual scram. 45% of'ated, and

3. If Region II of Figure 3.5.M-l b. Following any decrease is entered: of more than 10/ rated core flow while initial thermal power is greater than 40/ of rated.

BFH 3.5/4.5-20 Unit 2

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SURVEILLANCE REQUIREMENTS 3.5 Core and Containment Coolin S stems 4.5 Core and Containment Coolin S stems 3.5.M.3. (Cont'd)

Immediately initiate action and exit the region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by inserting control rods or by increasing core flow (starting a recirculation pump to exit the region is not an appropriate action), and

b. While exiting the region, immediately initiate a manual scram if thermal-hydraulic instability is observed, as evidenced by APRM oscillations which exceed 10 percent peak-to-peak of rated or LPRM oscillations which exceed 30 percent peak-to-peak of scale. If periodic LPRM upscale or downscale alarms occur, immediately check the APRM's and individual LPRM's for evidence of thermal-hydraulic instability.

BFN 3.5/4.5-20a Unit 2

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3.5 'BASES (Cont'd)

The minimum margin to the onset, of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1. A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit.

Because the probability of thermal-hydraulic oscillations is lower and the margin to the MCPR safety limit is greater in Region II than in Region I of figure 3.5.M-1, an immediate scram upon entry into the region is not necessary. However, in order to minimize the probability of core instability following entry into Region II, the operator will take immediate action to exit the region. Although formal surveillances are not performed while exiting Region lI (delaying exit for surveillances is undesirable), an immediate manual scram will be initiated instability is observed.

if evidence of thermal-hydraulic Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations). Periodic LPRM upscale or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated.

During regional oscillations, the safety limit MCPR is not approached until APRM oscillations are 30 percent peak-to-peak or larger in magnitude. In addition, periodic upscale or downscale LPRM alarms will occur before regional oscillations are large enough to threaten the MCPR safety limit. Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that core oscillations initiate while exiting Region II.

Normal operation of the reactor is restricted to thermal power and -core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.

3.5.N. References

1. Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 2, NEDO 24088-1 and Addenda.

2. "BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

3. Generic Reload Fuel Application, Licensing Topical Report, NEDE 24011-P-A and Addenda.

BFN Unit 2 3.5/4.5-32

4 PRIMAR RY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.6.E. J~et Pum s

2. Whenever there is recirculation flow with the reactor- in the STARTUP or RUN Mode and one recirculation pump is operating with the equalizer valve closed, the diffuser to lower plenum differential pressure shall be checked daily and the differential pressure of an individual jet pump in a loop shall not vary from the mean of all jet pump differential pressures in that loop by more than 10%.

3.6.F ecirculation um 0 e atio 4.6.F. Re i culation Pum 0 eration The reactor shall not be operated Recirculation pump speeds with one recirculation loop out shall be checked and logged of service for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. at least once per day.

With the reactor operating, if one recirculation loop is out of service, the plant shall be placed in a HOT SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is sooner returned 6o service.

2. Following one pump operation, 2. No additional surveillance the discharge valve of the low required.

speed pump may not be opened unless the speed of the faster pump is less than 50% of its rated speed.

3. When the reactor is not in the RUN mode, REACTOR POWER OPERATION with both 3. Before starting either recirculation pumps out-of-service recirculation pump for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted. during REACTOR POWER During such interval, restart of the OPERATION, check and recirculation pumps is permitted, lo'g the loop discharge provided the loop discharge temperature and dome temperature is within 75'F of saturation temperature.

the saturation temperature of BFN 3.6/4.6-12 Unit 2

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4 6 PRIMARY ARY

~ I l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6,F Rec rcu ation Pum 0 eratio 3.6.F.3 (Cont'd) the reactor vessel water as determined by dome pressure. The total elapsed time in natural circulation and one pump operation must be no greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.6.F.4 The reactor shall not be operated with both recirculation pumps out-of-service while the reactor is in the RUN mode. Following a trip of both recirculation pumps while in the RUN mode, immediately initiate a manual reactor scram.

3.6.G Structural Inte rit 4.6.G Structural Inte rit

1. The structural integrity of Inservice inspection of ASME ASME Code Class 1, 2, and Code Class 1, Class 2, and 3 equivalent components shall Class 3 components shall be be maintained in accordance performed in accordance with with Specification 4.6.G ,.Section XI 'of the ASME Boiler throughout the life of the and Pressure- Vessel Code and plant. applicable Addenda as required by 10 CFR 50, Section 50.55a(g),

a. With the structural except where specific written integrity of any relief has been granted by NRC Class 1 equivalent ASME'ode pursuant to 10 CFR 50, Section component, which is part 50.55a(g)(6)(i).

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not conforming to the above requirements, restore 2. Additional inspections the structural integrity of shall be performed on the affected component to certain circumferential within its limit or maintain pipe welds as listed to the reactor coolant system in provide additional either a Cold Shutdown condition protection against pipe or less than 50'F above the 'hip, which could damage minimum temperature required auxiliary and control by NDT considerations, until systems.

each indication of a defect has been investigated and evaluated.

3.6/4.6-13 BFN UNIT 2

~ '4 PRIMARY SYS EM BO ARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.G Structural Inte rit 4.6.G Structural Inte rit 3.6.G.1 (Cont'd) 4.6.G.2 ~Cont'd

b. With the structural integrity Feedwater GFW-9, KFW-13 of any ASME Code Class 2 or 3 GFW-12, GFW-26, equivalent component not KFW-31, GFW-29, conforming to the above KFW-39, GFW-15, requirements, restore the KFW-38, and GFW-32 structural integrity of the Main st earn GMS-6, affected component to zithin KMS-24 GMS-32, KMS-104 its limit or isolate the affected GMS-15, and GMS-24 component from all OPERABLE systems. DSRHR-4, DSRHR-7, DSRHR-6 Core Spray TCS-407, TSC-423, TSCS-408, and TSC-424 Reactor Cleanup DSRWC-4, DSRWC-3 DSRWC-6, DSRWC-5 HPCI THPCI 70 THPCI 70A THPCI 71 THPCI, 72 3.6/4.6-14 BFN Unit 2

3.6/4.6 BASES

~ ~ ~ ~ 3,.6.E/4.6.E (Cont'd) resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).

If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured. This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate. Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.

A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.

3.6.F/4.6.F Recirculation Pum 0 erat o Operation without forced recirculation is permitted for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is not in the RUN mode. And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75'F. This reduces the positive reactivity insertion to an acceptably low value.

Requiring at least one recirculation pump to be operable while in the RUN mode provides protection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions.

Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.

The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.

The program reflects the built-in limitations of access to the reactor coolant systems.

It is intended that the required examinations and inspection be completed during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.

BFN 3.6/4.6-32 Unit 2'

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t ENCLOSURE 2 DESCRIPTION AND jUSTIFICATION BROWNS FERRY NUCLEAR PLANT (BFN)

Reason for Chan e BFN unit 2 technical specifications Sections 3.5/4.5-M are being added and section. 3.6.F is being revised to incorporate surveillance requirements and Limiting Conditions for Operation (LCO) for reactor core thermal-hydraulic stability. These changes are being proposed to support the BFN unit 2 fuel reload technical specification submitted August 26, 1988 (TS 254) and to also implement the requirements of NRC Bulletin 88-07, Supplement l.

Bnack Bound General Design Criteria (GDC) 12 requires that reactor power oscillations either be (1) prevented or (2) detected and suppressed. The stability licensing basis for U.S. Boiling Water Reactors (BWRs) has been either that oscillations will not occur in allowable operating regions (as demonstrated by decay ratio calculations) or that oscillations can be detected and suppressed by reactor operators before protection limits are exceeded. In the past, BFN demonstrated compliance with GDC 12 by performing decay ratio analyses for each reload core to show that core thermal-hydraulic oscillations would not occur in allowable reactor operating regions.

Recent instability events at LaSalle and Vermont Yankee have led to concerns relative to the capability of currently approved analytical methods to adequately predict when instabilities will occur. These events as well as analyses performed by General Electric (GE) and NRC contractors indicate that instabilities may occur which result in regional oscillations and local power peaking greater than previously analyzed for in-phase core oscillations. In addition, preliminary calculations performed by GE indicate that under some operating conditions, the safety limit Minimum Critical Power Ratio (MCPR) may be violated during regional power oscillations.

The BWR Owners Group (BWROG) is curiently working with GE and NRC to develop a long-term resolution to stability concerns. In November 1988, interim corrective actions to address stability concerns were issued by GE and subsequently adopted by the BWROG. NRC Bulletin 88-07, Supplement 1

~December 30, 1988), requires ell BEE licensees to implement the GE recommendations. In addition, the bulletin requires some BWRs (BFN included) to initiate a manual scram following two recirculation pump trips when the reactor is in the RUN mode.

All BWR licensees have or are currently revising procedures to implement the requirements of the NRC bulletin. In discussions with BFN Site Licensing, NRC has indicated that in addition to procedural changes, BFN technical specifications must be modified to address stability concerns before restart of BFN unit 2.

Page 2 of 8 Descri tion and Justification for the P"o osed Chan e

l. Add LCO 3.5.M to read as follows:

Core Thermal-Hydraulic Stability

1. The reactor shall not be operated at a thermal power and core flow inside of Regions I and II of Figure 3.5.M-1.

2. If Region I of Figure 3.5.M-1 is entered, immediately initiate a manual scram.

3. If Region II of Figure 3.5.M-1 is entered:

a. Immediately initiate action and exit the region within 2 hours by inserting controls rods or by increasing core flow (starting a recirculation pump to exit the region is not an appropriate action), and

b. While exiting the region, immediately initiate a manual scram if thermal-hydraulic instability is observed, as evidenced by APRM oscillations which exceed 10 percent peak-to-peak of rated or LPRM oscillations which exceed 30 percent peak-to-peak of scale. If periodic LPRM upscale or downscale alarms occur, immediately check the APRM's and individual LPRM's for evidence of thermal-hydraulic instability.

Justification for Pro osed Chan e . .M The NRC bulletin (and the GE interim corrective actions), define regions of concern on the power/flow map referred=to as Regions A, B, and C. Region A includes operating conditions above the 100-percent rod line with core flow less than 40 percent of rated flow. Region B includes operating conditions between the 80- and 100-percent rod lines with core flow less than 40 percent of rated flow. Region C includes operating conditions above the 80-percent rod line with core flow between 40 and 45 percent of rated flow. The regions defined by GE cover the high power/low flow corner of the operating domain where stability margins are the lowest. The GE recommended region boundaries are based on plant operating experience, special stability tests, and analytical studies.

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~ ~ Justification or Proposed Chan e .5.M (Cont'd)

Region I of the proposed technical specification change corresponds to Region A as defined in the GZ interim corrective actions. Most oscillations have occurred during testing and operation at or above the 100-percent rod line with core flow near natural circulation. This behavior is consistent, with analyses which predict reduced stability margin with increasing power or decreasing flow. Region I bounds the majority of events and tests where core oscillations have been observed in GZ BWRs. This region represents the least stable conditions on the map and is therefore considered an excluded region in which normal power/'low operation is not allowed. Because operating experience has demonstrated that oscillations may rapidly develop in this region, operator actions are required to prevent the initiation of core oscillations in the event Region I is entered.

Region II of the proposed technical specification change includes both Region B and C defined in the GZ interim corrective actions. Region B of the interim corrective actions is also considered to be an excluded region (i.e., no intentionaI: entry) because of the relatively low core flow.

Zven though the probability of core oscillations is lover in Region B than in Region A, several events and tests have demonstrated that oscillations can occur in this region f'r certain operating conditions. However, because the power level is lower, the margin to fuel safety limits is greater in Region B. Region C of the interim corrective ac"ions i d efined as a buffer zone to the excluded reg'ons. Although no oscillations have been reported in this region, the possibility of core oscillations in this region cannot be ruled out. Therefore, the interim corrective actions allow operation in Region C only for control rod withdrawals during startup requiring fuel preconditioning. When operation does occur in Region C, the operator should'be aware of the possibility of core oscillations and procedures should ensure that adequate surveillance of nuclear instrumentation is performed. The proposed BFN technical specification change wil'1 combine Regions B and C into Region II and will conservatively apply Region B restrictions to Region C.

The potential for core thermal-hydraulic oscillations to occur when operating outside of Regions I and II is very small and therefore special restrictions are not required outside of these regions.

The region boundaries for the interim corrective actions were developed based on plant operating and test expe ience and analysis of GF. fuel designs. The regions were chosen to generically apply to all licensed GZ fuel designs and operating domains (e.g., Extended Load Line, Single Loop Operations). The BFif unit 2, cycle 6 core will contain four Westinghouse

('4) QUAD+ lead test assemblies. The presence of four QUAD+ bundles (about 0.5 percent of the core) will not significantly affect'he thermal-hydraulic stability characteristics of the core. In addition, the QUAD+ fuel design contains several design features which make the bundle more stable than the GZ 8 by 8 fuel design (reference 1). Therefore, the regions identified in the GZ interim corrective actions are appropriate for QUAD+ operation in BFN unit 2, cycle 6.

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Justi 'cat~on Eor P ooosed Chan e .5.M (Cont'd)

Proposed change 3.5.M.l restricts normal operation within guidelines of the power-flow map to conditions outside of Regions I and II. The excluded regions represent the least stable conditions for the plant.

Regions I and II are usually entered as a result of plant transients (recirculation pump trip) and are not part, of the normal- operating domain. All events (including test experience) that have resulted in core oscillations have occurred in either Region I or II. Intentional operation is not allowed in these regions in order to minimize the probability of encountering core oscillations and potentially challenging fuel safety limits.

Proposed change 3.5.M.2 requires the operator to manually scram the reactor if Region I is entered. Because BFN does not have a flow-biased unfiltered neutron flux scram, automatic scram protection is not provided until APRM oscillations reach a peak magnitude of 120 percent of rated power. Because of partial cancellation of out-of-phase LPRM signals during regional oscillations, local neutron flux can be significantly higher than indicated by the APRM signal. Preliminary calculations by GZ indicate that dur'ng operation in Region I, the safety limit MC?R (SLl<C?R) may be violated in some situations when APRM oscillations are approximately 15-45 percent peak-to-peak (reference 2). Because stabilit margins are the lowest in Region I, the potential exists for oscillations to rapidly increase in magnitude once they initiate. During transients which cause entry into Region I, the operator may not have sufficient time to manually insert control rods or increase core flow to suppress oscillations before they reach an unacceptable magnitude. The prompt action of manually scramming the reactor if Region I is entered will ensure adequate protection of the SLMCPR.

Proposed change 3.5.M.3 requires the operator to take immediate action to ex't Region II if entered. inadvertently. Because core thermal-hydraulic stability is very sensitive to core power and flow, stability margins are greater in Region II than in -Region I. The increased margin means that the probability of core oscillations is less and that oscillations will not increase in magnitude as rapidly as in Region I. Also, because of the lower power and/or increased core flow in Region II, the margin to the SLMCPR will be larger than in Region I. Because of the increased stability and SIMC?R margins, the operator will have more time to suppress oscillation in Region II before the SL4fCPR is violated.

Tests and operat.'ng experience have demonstrated that the insertion of control rods or the increase of core flow will rapidly dampen core thermal-hydraulic oscillations and move the plant into a region of increased stability margin. At reactor conditions where core oscillations begin, the insertion of a few control rod notches or a 1-2 percent increase in core flow will effectively suppress the oscillations (reference 2). Shen control rod insertion is used to exit the region, a predefined set of control rods will generally be used by the operator to ensure an expedient reduction in core thermal power. If one or more recirculation pumps are operational, increasing core flow is an acceptable alternative to inserting control rods and is generally simpler to perform. However, starting a recirculation pump to exit the region is not an appropriate action since it can lead to sudden reactivity insertions and initiate core oscillations. Also, starting a circulation pump can potentially distract the operators attention away fxom the detection of potential oscillations while in the region.

Justification for Pro osed Chan e .5.M (Cont'd)

The actions described above will minimize the probability of core thermal-hydraulic oscillations occurring following a transient which places the plant in Region II.

The presence of core thermal-hydraulic oscillations is an indication of the loss of control of the reactor, and if not mitigated, might rapidly lead to conditions which violate the SLMCPR. Experience has shown that oscillations can grow rapidly to high levels. Even though most events have been terminated by operator actions, it cannot be demonstrated that a manual control rod insertion or a core flow increase will always be rapid enough to prevent exceeding the SLMCPR. Therefore, if oscillations are detected, a rapid power reduction (i.e., scram) is the app'ropriate method of mitigation. This will ensure that oscillations are rapidly suppressed.

To avoid unnecessary challenges to the reactor protection system, a scram should be initiated only when there is clear evidence of core oscillations. APRM oscillations which exceed 10 percent peak-to-peak are clear indications of core thermal hydraulic instability. LPRM oscillations which exceed 30 percent peak-to-peak are approximately equivalent to APRM oscillations of 10 percent peak-to-peak during regional oscillations (reference 2) and are also clear indications of core instability. Periodic upscale or downscale LPRM alarms may be indicators of core thermal hydraulic oscillations. However, LPRM alarms alone should not be used to initiate a reactor scram because they only provide an indirect indication of oscillations and may be indicators of other conditions or equipment failures. If any LPRM alarms are received, the APRM's and individual LPRM's should be iaunediately evaluated to confirm the presence of oscillations.

Based on GE analyses (reference 2), the SLMCPR is not approached during regional oscillations until APRM oscillations are greater 'than approximately 30 percent peak-. to-peak. In addition, periodic upscale or downscale LPRM alarms will occur before regional oscillations are large enough to threaten the MCPR safety limit (reference 2). Therefore, initiating a manual scram on evidence of core instability as described above is sufficient to ensure that the SLMCPR will not be violated.

2. Add surveillance requirement 4.5.M to read as follows:

Core Thermal-Hydraulic Stability

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1. Verify that the reactor is outside of Region I and II of Figure 3.5.M.1:

a. Following any increase of more than 5% rated thermal "power while initial core flow is less than 45% of rated, and

b. Following any decrease of more than 10% rated core flow while initial thermal power is greater than 40% of rated.

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Page 6 of 8 Just cation for Pro osed Chan e 4 M The above surveillance requirements are being added to verify that the Reactor is operating in the proper region (acceptable region) when reactor power is increased greater than 5% rated thermal power (RTP) with ini'tial reactor core flow less than 45% or a decrease of 10% core flow while initial RTP is greater than 40 percent.

3. Add figure 3.5.M-l (See attached)

Justification for Pro osed Chan e Fi ure M-1 Figure 3.5.M-1, "BFN Power/Flow Stability Regions," provides the operators a clear illustration as to what conditions of reactor core power versus core flow are acceptable or unacceptable. Based on this table and the appropriate LCO 3.5.M, an operator can identify what action needs to be implemented to exit an unacceptable region.

4. Add Bases 3.5.M to read as follows:

The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1. A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit.

Because the probability of thermal-hydraulic oscillations is lower and the margin to the MCPR safety limit is greater- in Region II than in Region I of figure 3.5.M-l, an immediate scram upon entry into the region is not necessary. However, in order to minimize the probability of core instabiiity following entry into Region II, the operator will take immediate action to exit the region. Although formal surveillances are not performed while exiting Region II (delaying exit for surveillances is undesirable), an immediate manual scram if will be initiated evidence of thermal-hydraulic instability is observed.

Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM osci3 lations of 10 percent during regional oscillations). Periodic LPRM upscale, or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated.

During regional oscillations, the safety limit MCPR is not approached until APRM oscillations are 30 percent peak-to-peak or larger in magnitude. In addition, periodic upscale or downscale LPRM alarms will occur before regional oscillations are large enough to threaten the MCPR safety limit. Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit wilLnot be violated in the event that core oscillations initiate while exiting Region II.

Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.

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Page 7 of 8 Justi ication for Bases Section M This section is being updated to provide consistency and provide additional information supporting the reasoning for adding Limiting Condition for Operation and Surveillance Instruction 3.5.M-and 4.5.M.

5. Existing LCO 3.6.F.3 reads:

Steady-state operation with both recirculation pumps out-of-service for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted. During such interval restart of the recirculation pumps is permitted, provided the loop discharge temperature is within 75'F of the saturation temperature of...

Change existing LCO 3.6.F.3 to read as follows:

"When the reactor is not in the RUN mode," REACTOR POWER OPERATION with both recirculation pumps out-of-service for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted. During such interval, restart of the recirculation pumps is permitted, provided the loop discharge temperature is within 75'F of the saturation temperature of...

Justification for Pro osed Chan e LCO F This change is being made to clarify that when the reactor mode switch is NOT in the RUN position, that both recirculation pumps may be out of service for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during REACTOR POWER OPERATION.

6. Existing SI 4.6.F.3 reads:

Before starting either recirculation pump during steady-state operation, check and log the loop discharge temperature and dome saturation temperature.

Change Existing SI 4.6.F.3 to read; Before starting either recirculation pump during REACTOR POWER OPERATION, check and log the loop 'discharge temperature and dome saturation temperature.

Justification for Pro osed Chan e 4 6 F The word "reactor" is replacing the existing wording steady-state.

This will provide consistency with the change made in LCO 3.6.F.3.

7. Add LCO 3.6.F.4 to read as follows:

"The reactor shall not be operated with both recirculation pumps out of service while the reactor is in the RUN mode. Following a trip of both recirculation pumps while in the RUN mode, immediately initiate a manual reactor scram."

AaE>1 0 ~ ~, Page 8 of 8 Justi icatio fo LCO F 4 Proposed change 3.6.F.4 requires the operator to manually scram the reactor Following a trip of both recirculation pumps when the reactor is in the RUN mode. This action was not part of the GE interim corrective action recommendations but is a requirement of the NRC Bulletin. The reactor will enter either Region I or II following a recirculation pump trip from above the 80 percent rod line. Requiring a manual trip immediately following the loss of both recirculation pumps adds additional conservatism to ensure that thermal-hydraulic oscillations do not occur.

8. Revise Bases Section 3.6.F/4.6.F as follows:

"Operation without forced recirculation is permitted for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is not. in the RUN mode." And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75'F. This reduces the positive reactivity insertion to an acceptable low value.

"Requiring at least one recirculation pump to be operable while in the RUN mode provides protection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions."

Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50/ of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.

Justification for Revisin Bases Sect on F 4 F This section is being updated to reflect the proposed change in LCO 3.6.F..4.

References

1. WCAP-10507, "QUAD+ Demonstration Assembly Report," Westinghouse Electric Corporation, March 1984.

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2. General Electric Report prepared for the Boiling Water Reactors Owner,'.s Group, "Fuel thermal Margin During Core Thermal Hydraulic Oscillations in a Boiling Water Reactor," March 1989.

3. Letter, D. N. Grace (BWROG) to A. Thadani (NRC), "

Subject:

NRC Bulletin 88-07 Supplement 1, Power Oscillations in Boiling Water Reactors," January 26, 1989.

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ENCLOSURE 3 DETERMINATXO OF NO SIGNIFICANT HAZARDS CONSID RATION BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 Descri tion of P o osed Technical S ecificat on Amendment The proposed amendment would change the BFN technical specifications for unit 2 only. This amendment will add technical specification 3.5/4.5.M, Figure 3.5.M-1, "BFN Power/Flow Stability Regions," update bases section 3.5.M, revise section 3.6.F.3, add section 3.6.F.4, and update bases section 3.6.F/4.6.F. These changes will implement the requirements of NRC Bulletin 88-07, Supplement 1. These changes will define the reactor core regions of operation which are acceptable or unacceptable and provide require actions needed to exit operating in an unacceptable region.

Basis for Pro osed No Si nificant Hazards Consideration Determi ation NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating license involves no significant hazards considerations operation of the facility in accordance with the proposed amendment would if not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

This change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Implementation of the proposed TS change decreases the probability of core thermal-hydraulic oscillations by precluding operating conditions where instabilities have occurred at other plants. In addition, the proposed change will provide additional assurance that core oscillations that,do occur will be suppressed prior to exceeding fuel integrity limits. The proposed change does not have an adverse safety effect on any affected-safety system nor are the assumptions of the safety analyses affected by restricting operation to outside of Regions I and II. Therefore, the proposed change reduces the probability and consequences of potential core oscillations and does not increase the probability or consequences of any other previously analyzed event.

(2) This proposed change does not create the possibility of a new or different kind of accident from any previously analyzed. Restricting operation to outside of Regions I and II does not create any new failure mechanisms.

Plant procedures currently preclude normal operation in those regions.

Emergency entry into a restricted region is permitted to protect plant safety equipment provided that the prescribed actions (i.e., scram or

• exit) for the region entered are performed. Operator actions to exit Region II will be performed in compliance with all plant procedures, fuel preconditioning restrictions, and technical specifications.

(3) This change does not involve a significant reduction in a margin to safety. The proposed changes are conservative in nature and provide increased assurance that the fuel safety limit MCPR will not be violated due to core oscillations. These changes are consistent with NRC and GE guidelines. The implementation of this tech spec will actually increase this margin of safety at BFN by not allowing the plant to operate in Regions I or II. If one of these Regions are entered specific operator actions are required which will place the plant in a more conservative and safe condition than current BFN Tech Specs required.