ML20101F841

Proposed Tech Specs,Implementing BWROG Option I-D long-term Solution for Thermal Hydraulic Stability

ML20101F841
Person / Time
Site:	FitzPatrick
Issue date:	03/22/1996
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20101F830	List: JPN-96-009, Forwards Application for Amend to License DPR-59,revising TS to Implement BWROG Option I-D long-term Solution for Thermal Hydraulic Stability ML20101F831 ML20101F841
References
NUDOCS 9603260215
Download: ML20101F841 (40)

Text

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e Attachment I to JPN-96-009 REVISED TECHNICAL SPECIFICATION PAGES PROPOSED TECHNICAL SPECIFICATION CHANGES

, REGARDING IMPLEMENTATION OF BWROG OPTION 1-D l LONG-TERM SOLUTION FOR THERMAL HYDRAULIC STABILITY l

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New York Power Authority I

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 j DPR-59 9603260215 960322 PDR ADOCK 05000333 P PDR

LIST OF PAGE CHANGES Implementation of BWROG Option 1 D Long-Term Solution for Thermal Hydraulic Stability (JPTS-96-005)

Revise Appendix A as follows:

Removed Pages Insert Pages i

il li i vii vii 17 17 18 18 40 40 43a 43a l 124a 124a l 124b Delete 124c Delete 131 131 134 134 254: 254c 254d 254d  ;

254e 254e,f 1 254f 254e,f I l

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MBis_OF CONTENTS (Cont'd)

Eaaft :

F. ECCS-Cold Condition F. 122  !

G. Maintenance of Filled Discharge Nps G. 122a H. Average Planar Linear Heat Generation Rate H. 123 l

(APLHGR) l l. Linear Heat Generation Rate (LHGR) 1. 124 J. Thermal Hydraulic Stability DELETED 124a K. Single-Loop Operation NONE 124a SURVEILLANCE l LIMITING CONDITIONS FOR OPERATION REQUIREMENTS l

l 3.6 Reactor Coolant System 4.6 136 A. Pressurization and Thermal Limits A. 136 B. DELETED C. Coolant Chemistry C. 139 D. Coolant Leakage D. 141 l E. Safety and SMety/ Relief Valves E. 142a F. Structural integrity F. 144 i l

G. Jet Pumps G. 144 l l

H. DELETED

l. Shock Suppressors (Snubbers) 1. 145b l 3.7 Containment Systems 4.7 165 l A. Primary Containment A. 165 l

B. Standby Gas Treatment System B. 181  !

! C. Secondary Containment C. 184 l i

D. Primary Containment Isolation Valves D. 185 l

! l l 3.8 Miscellaneous Radioactive Material Sources 4.8 214 l

3.9 Auxiliary Electrical Systems 4.9 215  ;

. l i A. Normal and Reserve AC Power Systems A. 215 l l B. Emergency AC Power System B. 216 )

l C. Diesel Fuel C. 218 l D. Diesel-Generator Operability D. 220 E. Station Batteries E. 221 F. LPCI MOV Independent Power Supplies F. 222a

G. Reactor Protection System Electrical Protection G. 222c Assemblies 3.10 Core Alterations 4.10 227

, A. Refueling Interlocks A. 227 l B. Core Monitoring B. 230 C. Spent Fuel Storage Pool Water Level C. 231 D. Control Rod and Control Rod Drive Maintenance D. 231 3.11 Additional Safety Related Plant Capabilities 4.11 237 A. Main Control Room Ventilation A. 237 B. Crescent Area Ventilation B. 239 C. Battery Room Ventilation C. 239 I

I Amendment No. GO, 13,98,113,1SS, ii

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0 JAFNPP LIST OF FIGURES l

Fiaures lit!A EaQA 4.1-1 (Deleted) 4.2-1 (Deleted) >

3.4-1 Sodium Pentaborate Solution (Minimum 34.7 B-10 Atom % Enriched) 110 Volume-Concentration Requirements

( 3.4-2 Saturation Temperature of Enriched Sodium Pentaborate Solution 111 l

l 3.5-1 (Deleted) 1 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 12 EFPY 163 Part 1 l

l 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 14 EFPY 163a

Part 2 l

t 3.6-1 Reactor Vessel Pressura - Temperature Limits Through 16 EFPY 163b Part 3 4.6-1 Chloride Stress Corrosion Test Results at 500*F 164 6.1-1 (Deleted) l 6.2-1 (Deleted) l l

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Amendment No. l i, 22, 4 2, Si, 72, 74, SS, 00,109,113, ' 1 S, l ' 7,131,137,153,1 S2, 227, ,

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2.1 BASES (Cont'd) power, the rate of power rise is very slow. Generally, the in order to ensure that the IRM provided adequate heat flux is in near equilibrium with the fission rate. In an protection against the single rod withdrawal error, a range assumed uniform rod withdrawal approach to the scram of rod withdrawal accidents was analyzed. This analysis level, the rate of power rise is no more than 5 percent of included starting the accident at various power levels. The rated power per minute, and the APRM system would be most severe case involves an initial condition in which the more than adequate to assure a scram before the power reactor is just subcritical and the IRM system is not yet on could exceed the safety limit. The 15 percent APRM scale. This condition exists at quarter rod density. scram remains active until the mode switch is placed in the Additional conservatism was taken in this analysis by RUN position. This switch occurs when reactor pressure is assuming that the IRM channel closest to the withdrawn greater than 850 psig.

rod is by-passed. The results of this analysis show that the reactor is scrammed and peak power limited to one percent c. APRM Flux Scram Trio Settina (Run Mode) of rated power, thus maintaining MCPR above the Safety Limit. Based on the above analysis, the IRM provides The APRM system obtains neutron flux input signals from protection against local control rod withdrawal errors and LPRMs (fission chambers) and is calibrated to indicate

, continuous withdrawal of control rods in sequence and percent rated thermal power. The APRM scrams in the run provides backup protection for the APRM. mode are a flow referenced scram and a fixed high neutron flux scram. As power rises during transients, the

b. APRM Flux Scram Trio Settina (Refuel or Startuo and Hot instantaneous neutron flux (as a percentage of rated) will Standbv Mode) rise faster than the rate of heat transfer from the fuel (percentage of rated thermal power) due to the thermal For operation in the startup mode while the reactor is at time constant of the fuel and core thermal power will be low pressure, the APRM scram setting of 15 percent of less than the power indicated by the APRMs (neutron flux) rated power provides adequate thermal margin between at either scram setting.

the setpoint and the safety limit,25 percent of rated. The margin is adequate to accommodate anticipated maneuvers The APRM flow referenced scram trip setting, nominally associated with power plant startup. Effects of increasing varies from 54% power at 0% recirculation flow to 120%

pressure at zero or low void content are minor, cold water power at 100% recirculation flow but is limited to 117%

from sources available during startup is not much colder rated power. The flow referenced trip will result in a than that already in the system, temperature coefficients significantly earlier scram during slow thermal transients, are small, and control rod pattems are constrained to be such as the loss of 80'F feedwater heating event, than uniform by operating procedures backed up by the rod would result from the 120% fixed high neutron flux scram.

worth minimizer and the Rod Sequence Control System. The lower flow referenced scram setpoint therefore Worth of individual rods is very low in a uniform rod decreases the severity (aCPR) of a slow thermal transient pattern. Thus, of all possible sources of reactivity input, and allows lower MCPR Operating Limits if such a transient uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated Amendment No. 43, 17

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2.1 BASES (Cont'd)

c. APRM Flux Scram Trio Settina (Run Mode) (cont'd) APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant is the limiting abnormal operational transient during a recirculation flow rate, and thus provides an added

• certain exposure interval in the cycle. The flow level of protection before APRM Scram. This rod referenced trip also provides protection for power block trip setting, which is automatically varied with i oscillations which may result from reactor thermal recirculation loop flow rate, prevents an increase in -

hydraulic instability. the reactor power level to excessive values due to control withdrawal. The flow variable trip setting l The APRM fixed high neutron flux scram protects the parallels that of the APRM Scram and provides  ;

reactor during fast power increase transients if credit margin to scram, assuming a steady-state operation is not taken for a direct (position) scram or flow at the trip setting, over the entire recirculation flow ,

referenced scram. range. The actual power distribution in the core is '

established by specified control rod sequences and is  :

The scram trip setting must be adjusted to ensure monitored continuously by the in-core LPRM system.  ;

that the LHGR transient peak is not increased for any As with the APRM scram trip setting, the APRM rod combination of maximum fraction of limiting power block trip setting is adjusted downward if the density (MFLPD) and reactor core thermal power. maximum fraction of limiting power density exceeds The scram setting is adjusted as specified in Table the fraction of rated power, thus preserving the  !

3.1-1 when the MFLPD is greater than the fraction of APRM rod block margin. As with the scram setting, rated power (FRP). This adjustment may be this may be accomplished by adjusting the APRM '

accomplished by either reducing the APRM scram and gain.

rod block settings or adjusting the indicated APRM  ;

signal to reflect the high peaking condition. 2. Reactor Water Low Level Scram Trio Settina Analyses of the limiting transients show that no The reactor low water level scram is set at a point which scram adjustment is required to assure that the will assure that the water levet used in the Bases for the MCPR will be greater than the Safety Limit when the Safety Limit is maintained. The scram setpoint is based on j transient is initiated from the MCPR operating limits normal operating temperature and pressure conditions specified in the Core Operating Limits Report. because the level instrumentation is density compensated. I

d. APRM Rod Block Trio Settina Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The Amendment No. 'P, '19,1S2, 18

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JAFNPP TABLE 3.1-1 REACTOR PROTECTION SYSTEM (SCRAMI INSTRUMENTATION REQUIREMENTS ,

Minimum No. of Mode in Which Function Operable Instrument Must Be Operable Total Number of Channels Per Instrument Channels Trip System Refuel Startup Run Provided by Design ,

(Notes 1 and 2) Trip Function Trip Level Setting (Note 7) for Both Trip Systems Action (Note 3) ,

t 1 Mode Switch X X X 1 Mode Switch A in Shutdown ,

1 Manual Scram X X X 2 A i

3 IRM High Flux s96% (120/125) X X 8 A ,

of full scale 3 IRM Inoperative X X 8 A 2 APRM Neutron Flux- s15% Power X X 6 A Startup (Note 15) 2 APRM Flow Referenced (Note 12) X 6 A or B Neutron Flux (Not to l exceed 117%) (Note 13) .

2 APRM Fixed High s120% Power X 6 A or B Neutron Flux 2 APRM Inoperative (Note 10) X 'X X 6 A or B Amendment No. 'i, 18,183,227, 40 t

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TABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS

12. The APRM Flow Referenced Neutron Flux Scram setting shall be less than or equal to the limit specified in the Core Operating Limits Report.

13. The Average Power Range Monitor scram function is varied as a function of recirculation flow (W). The trip setting of this function must be maintained as specified in the Core Operating Limits Report. t

14. Deleted.

15. This Average Power Range Monitor scram function is fixed point and is increased when the reactor mode switch is placed in the Run position.

16. Instrumentation common to PCIS.

Amendment No. 183,227, 43a

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3.5 (cont'd)

J. Thermal Hydraulic Stability

1. When the reactor is in the run mode: ,

a. Under normal operating conditions the reactor shall not be intentionally operated within the Power / Flow Exclusion Region defined in the Core Operating Limits Report (COLR).

b. If the reactor has entered the Power / Flow Exclusion Region, the operator shall immediately insert control rods and/or increase recirculation flow to establish operation outside the region.

K. Sinale-Loon Ooeration

1. The reactor may be started and operated, or reactor operation may continue, with a single Reactor Coolant System recirculation loop in operation. The requirements applicable to single-loop operation in Specifications 1.1.A, 2.1.A, 3.1.A, 3.1.B,3.2.C, and 3.5.H shall be in effect within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or the reactor shall be placed in at least the hot shutdown mode within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. During resumption of two-loop operation following a period of single-loop operation, the discharge valve of the lower speed pump shall not be opened unless the speed of the faster pump is less than 50 percent of its rated speed.

3. With no Reactor Coolant System recirculation loop in service, the reactor shall be placed in at least the hot shutdown mode within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• Amendment No. 39r98, 124a

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2 3.5 BASES (cont'd)

J. Thermal Hydraulic Stability K. Sinale-Loon Operation 10 CFR 50, Appendix A, General Design Criterion 12 Requiring the discharge valve of the lower speed loop to requires that power oscillations are either prevented or can remain closed until the speed of the faster pump is below be readily detected and suppressed without exceeding 50 percent of its rated speed provides assurance when specified fuel design limits. To minimize the likelihood of a going from one to two pump operation that excessive thermal hydraulic instability which results in power vibration of the jet pump risers will not occur.

oscillations, a power / flow exclusion region to be avoided during normal operation is calculated using the approved L. References methodology specified in Technical Specification 6.9(A)4.

Since the exclusion region may change each fuel cycle, the 1. "FitzPatrick Nuclear Power Plant Single-Loop limits are contained in the Core Operating Limits Report. Operation", NEDO-24281, August 1980.

Specific directions are provided to avoid operation in the exclusion region and to immediately exit the region if 2. " Application of the ' Regional Exclusion with Flow-entered. Entries into the exclusion region are not part of Biased APRM Neutron Flux Scram' Stability Solution normal operation, but may result from an abnormal event, (Option 1-D) to the James A. FitzPatrick Nuclear such as a single recirculation pump trip or loss of Power Plant," GENE-637-044-0295, February 1995 feedwater heating, or be required to prevent equipment damage. In these events, time spent within the exclusion region is minimized.

Although operator actions can prevent the occurrence of and protect the reactor from an instability, the APRM flow-biased reactor scram will suppress power oscillations prior l to exceeding the fuel safety limit (MCPR). Reference i 3.5.L.2 demonstrated that this protection is provided at a  !

high statistical confidence level for core-wide mode  ;

oscillations and at a nominal statistical confidence level for i regional mode oscillations. This reference also demonstrated that the core-wide mode of oscillation is preferred due to the large single-phase channel pressure drop associated with the small fuel inlet orifice diameters.

t Amendment No. 'i, Si, 90, j 131

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(THIS PAGE INTENTIONALLY BLANK) l Amendment No. 44r30, 52, S', 98, 134

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! JAFNPP l (A) ROUTINE REPORTS (Continued) l l

l 4. CORE OPERATING LIMITS REPORT l

l a. Core operating limits shall be established prior to startup from each reload l cycle, or prior to any remaining portion of a reload cycle for the following: l l

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• The Average Planar Linear Heat Generation Rates (APLHGR) of Specification 3.5.H;

• The Minimum Critical Power Ratio (MCPR) and MCPR low flow adjustment factor, K,, of Specifications 3.1.8 and 4.1.E; l

• The Linear Heat Generation Rate (LHGR) of Specification 3.5.l; l

• The Reactor Protection System (RPS) APRM flow biased trip settings of Table 3.1-1; I l

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• The flow biased APRM and Rod Block Monitor (RBM) rod block settings of i

Table 3.2-3; and l l

• The Power / Flow Exclusion Region of Specification 3.5.J. I and shall be documented in the Core Operating Limits Report (COLR).

. b. The analytical methods used to determine the core operating limits shall be I those previously reviewed and approved by the NRC as described in:

1. " General Electric Standard Application for Reactor Fuel," NEDE-24011-P, latest approved version and amendments.

2. " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR - LOCA Loss-of-Coolant Accident Analysis," NEDC-31317P, October,1986 including latest errata and addenda, l i

3. " Loss-of-Coolant Accident Analysis for James A. FitzPatrick Nuclear Power Plant," NEDO-21662-2, July,1977 including latest errata and i addenda.

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4. "BWR Owners' Group Long-term Stability Solutions Licensing Methodology," NEDO-31960-A, June 1991.

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5. "BWR Owners' Group Long-term Stability Solutions Licensing Methodology," NEDO-31960-A, Supplement 1, March 1992.

1 Amendment No. 442, 254c

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c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydrau' _ in ':ts, ECCS limits, nuclear limits such as shutdown margin, and transient ano ' c.:ident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements thereto, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.

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Amendment No. 32,110,162, 254d I

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. Amendment No. 32,110,1S2, 254e,f l

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l l Attachment ll to JPN-96-009 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING IMPLEMENTATION OF BWROG OPTION 1-D LONG-TERM SOLUTION FOR THERMAL HYDRAULIC STABILITY i

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333

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Attachm:nt ll to JPN-96-009 SAFETY EVALUATION l Page 1 of 8 l

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1. DESCRIPTION OF THE PROPOSED CHANGES The following proposed changes to the James A. FitzPatrick Technical Specifications l

establish operability requirements for avoidance and protection from thermal hydraulic instabilities to be consistent with BWROG long-term solution Option I-D (References 1

! and 2). Editorial changes are also made to support the revised specifications, improve readability of Bases sections, and enhance the presentation of requirements for single loop operation.

l l Reference 3 issued an amendment with similar Technical Specification changes to the l

Vermont Yankee Nuclear Power Station (VYNPS). VYNPS is the lead plant for the Option I-D stability solution.

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l Replace SR 3.5.J with " DELETED." Add line for LCO (3.5) "K. Single-Loop Operation"  !

with no associated SR, denoted by "NONE" on page 124a. '

l Paae vii Replace title for Figure 3.5-1 with "(Deleted)," delete page number.

Paaes 17 and 18 Change BASES Section 2.1.c to read:

"The APRM system obtains neutron flux input signals from LPRMs (fission chambers)

and is calibrated to indicate percent rated thermal power. The APRM scrams in the l

run mode are a flow referenced scram and a fixed high neutron flux scram. As power rises during transients, the instantaneous neutron flux (as a percentage of rated) will

! rise faster than the rate of heat transfer from the fuel (percentage of rated thermal l power) due to the thermal time constant of the fuel and core thermal power will be less than the power indicated by the APRMs (neutron flux) at either scram setting.

The APRM flow referenced scram trip setting, nominally varies from 54% power at 0%

l recirculation flow to 120% power at 100% recirculation flow but is limited to 117% rated l power. The flow referenced trip will result in a significantly earlier scram during slow thermal transients, such as the loss of 80 F feedwater heating event, than would result from the 120% fixed high neutron flux scram. The lower flow referenced scram

, setpoint therefore decreases the severity (aCPR) of a slow thermal transient and allows L

lower MCPR Operating Limits if such a transient is the limiting abnormal operational transient during a certain exposure intervalin the cycle. The flow referenced trip also provides protection for power oscillations which may result from reactor thermal hydraulic instability.

The APRM fixed high neutron flux scram protects the reactor during fast power increase transients if credit is not taken for direct (position) scram or flow referenced i scram, i l

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Attachment 11 t2 JPN-96-009 SAFETY EVALUATION Page 2 of 8 i

increase transients if credit is not taken for direct (position) scram or flow referenced scram.

The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of maximum fraction of limiting power density (MFLPD) and reactor core thermal power. The scram setting is adjusted as specified in Table

3.1-1 when the MFLPD is greater than the fraction of rated power (FRP). This adjustment may be accomplished by either reducing the APRM scram and rod block settings or adjusting the indicated APRM signal to reflect the high peaking condition.

Analyses of the limiting transients show that no scram adjustment is required to assure that the MCPR will be greater than the Safety Limit when the transient is initiated from the MCPR operating limits specified in the Core Operating Limits Report."

Paae 40 Delete reference to Note 14 for APRM Flow Referenced Neutron Flux. The revised l specification reads, "APRM Flow Referenced Neutron Flux (Not to exceed 117%) (Note

13)."

Delete reference to Note 14 for APRM Fixed High Neutron Flux. The revised specification reads, "APRM Fixed High Neutron Flux."

l Paae 43a l Delete text for Note 14 and replace with " Deleted."

l l Paaes 124a throuah 124c l Change LCO 3.5.J to read as follows:

"J. Thermal Hydraulic Stability l 1. When the reactor is in the run mode:

a. Under normal operating conditions the reactor shall not be intentionally operated within the Power / Flow Exclusion Region defined in the Core Operating Limits Report (COLR).

b. If the reactor has entered the Power / Flow Exclusion Region, the operator shall immediately insert control rods and/or increase

recirculation flow to establish operation outside the region."

Delete LCOs 3.5.J.2 and 3.5.J.3.

Delete SR 4.5.J.1.

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Attachmtnt 11 to JPN-96-009 SAFETY EVALUATION Page 3 of 8 Renumber and revise LCOs 3.5.J.4 through 3.5.J.6 to read:

"K. Sinole-Loon Ooeration 1

1. The reactor may be started and operated, or reactor operation may continue, with a single Reactor Coolant System recirculation loop in operation. The requirements applicable to single-loop operation in Specifications 1.1.A,2.1.A, 3.1.A,3.1.B,3.2.C, and 3.5.H shall be in effect within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or the reactor shall be placed in at least the hot shutdown mode within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. During resumption of two-loop operation following a period of single-loop ,

operation, the discharge valve of the lower speed pump shall not be opened '

unless the speed of the faster pump is less than 50 percent of its rated speed.

l 3. With no Reactor Coolant System recirculation loop in service, the reactor shall be placed in at least the hot shutdown mode within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

Delete pages 124b and c.

Paae 131 Revise Bases Section 3.5.J to read:

"J. Thermal Hydraulic Stability 10 CFR 50, Appendix A, General Design Criterion 12 requires that power oscillations are either prevented or can be readily detected and suppressed without exceeding specified fuel design limits. To minimize the likelihood of a thermal hydraulic instability which results in power oscillations, a power / flow exclusion r00i on to be avoided during normal operation is calculated using the approved methodology specified in Technical Specification 6.9(A)4. Since the exclusion region may change each fuel cycle, the limits are contained in the 1 Core Operating Limits Report. Specific directions are provided to avoid I operation in the exclusion region and to immediately exit the region if entered.

Entries into the exclusion region are not part of normal operation, but may result from an abnormal event, such as a single recirculation pump trip or loss of feedwater heating, or be required to prevent equipment damage. in these events, time spent within the exclusion region is minimized.

Although operator actions can prevent the occurrence of and protect the reactor -

from an instability, the APRM flow-biased reactor scram will suppress power oscillations prior to exceeding the fuel safety limit (MCPR). Reference 3.5.L.2 I demonstrated that this protection is provided at a high statistical confidence

level for core-wide mode oscillations and at a nominal statistical confidence level for regional mode oscillations. This reference also demonstrated that the ,

j core-wide mode of oscillation is preferred due to the large single-phase channel pressure drop associated with the small fuel inlet orifice diameters."-

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Attachment ll t2 JPN-96-009 SAFETY EVALUATION Page 4 of 8 Relocate the last paragraph of Bases Section 3.5.J to a revised Bases Section 3.5.K, renumber Bases Section 3.5.K to be Bases Section 3.5.L and add reference 3.5.L.2:

"K. Single-Looo Ooeration Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50 percent 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.

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1. "FitzPatrick Nuclear Power Plant Single-Loop Operation", NEDO-24281, August 1980.

2. " Application of the ' Regional Exclusion with Flow-Biased APRM Neutron Flux Scram' Stability Solution (Option I-D) to the James A. FitzPatrick Nuclear Power Plant," GENE-637-044-0295, February 1995."

Page 134 Delete Figure 3.5-1, mark page "(THIS PAGE INTENTIONALLY BLANK)."

Page 254-c through f Change page number 254-c to 254c. l Add another bullet to section 6.9(A)4.a:

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". The Power / Flow Exclusion Region of Specification 3.5.J."

I Add new sections 6.9(A)4.b.4 and 6.9(A)4.b.5:

l "4. "BWR Owners' Group Long-term Stability Solutions Licensing Methodology,"

NEDO-31960-A, June 1991.

5. "BWR Owners' Group Long-term Stability Solutions Licensing Methodology,"

NEDO-31960-A, Supplement 1, March 1992."

Move sections 6.9(A)4.c and 6.9(A)4.d to page 254d. Change page number of following page to "254e,f."

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Attachm:nt il to JPN-96-009 i SAFETY EVALUATION j Page 5 of 8 ll. PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed changes is to implement controls over plant operation which ensure compliance with General Design Criteria (GDC) 10 and 12 of 10 CFR 50 Appendix A.

Following the thermal hydraulic instability event at LaSalle in 1988 (Reference 4), a BWR Owners' Group committee was formed to obtain resolution of NRC concerns over compliance with GDC 10 and 12. The BWR Owners' Group developed several stability long-term solutions which addressed these issues, which were subsequently approved by the NRC (References 1 and 2).

NYPA has chosen to implement the Option I-D stability solution and has submitted a plant unique assessment demonstrating the suitability of the solution for FitzPatrick to the NRC (Reference 5). Option I-D requires use of the flow-biased Average Power Range Monitor high neutron flux scram without Simulated Thermal-Power Monitor (STPM), and a power / flow map exclusion region while adhering to certain power distribution limitations.

The purpose of changes related to removal of the STPM is to ensure the Technical Specifications reflect the plant configuration following modification of the APRMs to support the Option I-D stability solution.

The purpose of changes related to replacement of a generic restricted region (Figure 3.5-1, being removed from the Technical Specifications) which may be entered with appopriate monitoring, with a cycle specific exclusion region (to be provided in the COLR) which is not to be entered during normal operation, is to prevent the occurrence of thermal hydraulic oscillations while removing unnecessary operating limitations.

The purpose of the editorial changes separating the requirements for thermal hydraulic stability from the requirements for single loop operation is to enhance the clarity of the Technical Specifications.

The purpose of specifying establishing the thermal hydrauhc stability exclusion region on a cyclic basis in the COLR is to ensure that the region is correct for the reload core design.

Other editorial changes made are the addition or deletion of blank pages as required by expanded or contracted text sections and revision of the table of contents to reflect changed section numbering.

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Attachment ll t3 JPN-96-009 SAFETY EVALUATION Page 6 of 8 111. SAFETY IMPLICATIONS OF THE PROPOSED CHANGES Changes to pages 17,18,40 and 43a reflect removal of the STPM function from the APRM flow-biased scram trip circuit. This allows the flow referenced scram to provide protection of the fuel safety limit (minimum critical power ratio, MCPR) for reactor instabilities. It is necessary to disable the filter of the STPM (which models the thermal time constant of the fuel) so that the scram trip circuit will provide adequate protection for thermal hydraulic oscillations. The density wave oscillations associated with thermal hydraulic instability have a natural frequency of 0.3 to 0.7 Hz, which corresponds to a period of 1.4 to 3 seconds, compared to a time constant of about 6 seconds for the STPM. When the APRM neutron flux signalis processed through the STPM, the magnitude of the oscillations associated with thermal hydraulic instability will be greatly reduced because the STPM time constant is significantly longer than the period of the oscillation. With the STPM processing removed from the APRM flow referenced scram trip input, the signal will not be reduced in magnitude, and a scram will occur in time to limit the ACPR to a value such that the MCPR safety limit is not violated.

The analyses presented in Reference 5 demonstrate that the flow-biased APRM scram will provide suppression of thermal hydraulic oscillations prior to the MCPR safety limit being violated. These analyses apply to core-wide oscillations with a high degree of statistical confidence, and to regional oscillations with a nominal degree of statistical confidence.

Removal of the STPM filter may result in reactor scrams which would not have occurred with the time constant in place. In the event of such a scram, compliance with Technical Specification limitations on Reactor Coolant System (RCS) heatup and cooldown rates, and design limitations on RCS thermal cycles will ensure RCS thermal stresses are adequately controlled and accounted for. Because the unfiltered APRM flow biased trip provides protection for the MCPR safety limit for potential reactor instabilities, a net safety benefit is gained.

Changes to pages 124a through 124c,131 and 134 include those which provide increased protection against potential fuel damage as a result of unstable reactor operation. The present Technical Specifications allow operation within a generically defined region of the core power / flow map in which thermal hydraulic instability is thought to be possible, provided the operators utilize enhanced monitoring to detect an instability and provided action is taken to suppress an instability, should it occur. The proposed changes prohibit normal operation within a plant specific power / flow exclusion region in which instabilities are conservatively predicted to occur. If entry is made into the exclusion region as a result of a plant transient, the proposed changes require immediate exit by either increasing reactor water recirculation flow or by inserting control rods. Therefore, the proposed changes reduce the potential of occurrence of thermal hydraulic instabilities.

An editorial change was made to these pages to separate the controls required for protection against instability from those which are required to permit single loop operation. The controls for single loop operation were relocated to a separate section (3.5.K) of the Technical Specifications with no change in technical content. The power / flow exclusion region is not dependent on the number of recirculation loops in

Attachment 11 to JPN-96-009 SAFETY EVALUATION Page 7 of 8 l

service, therefore it is not necessary to couple the requirements for single loop operation with those related to thermal hydraulic stability. This change improves the clarity of the Technical Specifications.

The change to page 254c requires that the power / flow exclusion region be established for each operating cycle in accordance with NRC approved methods and included in the Core Operating Limits Report. This ensures that the stability limits are appropriate for the current core configuration.

Editorial changes to the table of contents and page additions and deletions do not alter any operability or surveillance requirements contained in the Technical Specifications.

Therefore, these changes have no effect on safety.

l IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION I l

Operation of the FitzPatrick plant in accordance with the proposed Amendment would I not involve a significant hazards consideration as defined in 10 CFR 50.92, since it I would not:

1. involve a significant increase in the probability or consequences of an accident previously evaluated because:

l The implementation of BWR Owners' Group long-term stability solution Option I-D at FitzPatrick does not modify the assumptions contained in the existing accident analysis. The use of an exclusion region and the operator actions required to avoid and minimize operation inside the region do not increase the l possibility of an accident. Conditions of operation outside of the exclusion region are within the analytical envelope of the existing safety analysis. The j operator action requirement to exit the exclusion region upon entry minimizes l the possibility of an oscillation occurring. The actions to drive control rods I and/or to increase recirculation flow to exit the region are maneuvers within the envelope of normal plant evolutions. The flow referenced scram has been analyzed and will provide automatic fuel protection in the event of an instability.

Thus, each proposed operating requirement provides defense in depth for i protection from an instability event while maintaining the existing assumptions of the accident analysis.

I 2. create the possibility of a new or different kind of accident from those previously evaluated because:

The proposed operating requirements either mandate operation within the envelope of existing plant operating conditions or force specific operating maneuvers within those carried out in normal operation. Since operation of the plant with all of the proposed requirements are within the existing operating basis, an unanalyzed accident will not be created through implementation of the proposed change.

3. involve a significant reduction in the margin of safety because:

Each of the proposed requirements for plant thermal hydraulic stability provides l

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Attachm:nt il to JPN-96-009 i

SAFETY EVALUATION l

l Page 8 of 8 l

4 a means for fuel protection. The combination of avoiding possible unstable conditions and the automatic flow referenced reactor scram provides an in depth means for fuel protection. Therefore, the individual or combination of  ;

means to avoid and suppress an instability supplements the margin of safety.

V. IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not adversely affect the ALARA or Fire l Protection Program at the FitzPatrick plant, nor will the changes impact the environment.

VI. CONCLUSION i

l Based on the discussions above, the implementation of the BWROG long term stability solution Option I-D at FitzPatrick does not involve a significant hazards censideration, or an unreviewed safety question, and will not endanger the health and safety of the public. The Plant Operating Review Committee and Safety Review Committee have i j reviewed this proposed Technical Specification change and agree with this conclusion.

Vll. REFERENCES (1) BWR Owners' Group Long-Term Stability Solutions Licensing Methodology, NEDO-31960-A, June 1991. i l

(2) BWR Owners' Group Long-Term Stability Solutions Licensing Methodology, NEDO-31960-A, Supplement 1, March 1992 (3) NRC Letter, Daniel H. Dorman to Donald A. Reid, " Issuance of Amendment (TAC NO. M89201)," dated August 9,1995 l (4) NRC Information Notice No. 88-39: LaSalle Unit 2 Loss of Recirculation Pumps i with Power Oscillation Event, June 15,1988 (5) NYPA Letter, William J. Cahill, Jr. to NRC (JPN-95-032), " Submittal of Plant Specific Licensing Topical Report for Long-Term Solution on Reactor Stability l (Generic Letter 94-02)," dated June 29,1995 l

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Attachment til to JPN 96-009 MARKUP OF TECHNICAL SPECIFICATION PAGES PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING IMPLEMENTATION OF BWROG OPTION 1-D LONG-TERM SOLUTION FOR THERMAL HYDRAULIC STABILITY l

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

1 . JAFNPP TABLE OF CONTENTS (Cont'd)

! F. ECCSCold Condition F.

s G. Maintenance of Filled Discharge Pipe l 1 G. 122a '

H.

I.

Avera0s Planar Unear Heat Generation Rate (APLHGR) H. 123 f Unear Heat Generation Rate (LHGR) 1. 124 i J. Thermal Hydraulic Stability ]

j K.

-d:6 ow'D 124a -

SM9e. g, oge<ch'^

SURVElbCE j UMITING CONDITIONS FOR OPERATION REQUIREMENTS i

! 3.6 Reactor Coolant System 5

4.8 136 A. Pressurization and Thermal Umits A. 136 i B. DEETED 4

C. Coolant Chemistry

} C. 139 D. Coolant Laakage D. 141

{ E. Safety and Safety / Relief Valves . E. 142a F. Structuralintegrity F. 144 G. Jet Pumps G. 144 H. DELETED

1. Shock Suppressors (Snubbers) l

1. 145b

3.7 Containment Systems

' 4.7 165

} A. Primary Containment A. 165 j . B. Standby Gas Treatment System '

B. 181 j C. Secondary Containment C. 184 l D. Primary Containment isolation Valves D. 185  ;

l 3.8 Miscellaneous Radioactive Material Sources 4.8 214 3.9 Auxiliary Electrical Systems l

4.9 215 I

A. Normaland Reserve AC Power Systems A. 215 i B.

Emergency AC Power System B. 216 j C. Diesel Fuel C. 218 l D. Diesel Generator Operability D. 220 i E. Station Batteries E. 221 i F. LPCI MOVIndependent Power Supplies F. 222a l G. Reactor Protection System Electrical Protection Assemblies G. 222c i 3.10 Core Alterations 4.10 227 i A. RefuelingInterlocks A. 227

! B. Core Monitoring B. 230

? C. Spent Fuel Storage PoolWater Level C. 231 j D. Control Rod and Control Rod Drive Maintenance D. 231

. 3.11 Additional Safety Related Plant Capabilities 4.11 237

A. Main Control Room Ventilation A. 237 i B. Crescent Area Ventilation B. 239

C. Battery Room Ventilation C. 239 i

e Amendment No. f, f, %, if +60, I

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JAFNPP LIST OF FIGURES Fiauras I,,t[a EA21 4.1-1 (Deleted) 4.2-1 (Deleted) 3.4-1 Sodium Pentaborate Solution (Minimum 34.7 B-10 Atom % Enriched) 110 Volume Concentration Requirements 3.4-2 Saturation Temperature of Enriched Sodium Pentaborate Solution 111 3.5-1 *ar.T,e " ewer e..J Oeie .9ew U.TJ:e of Os,ee:::ee'Jee. 3.5.J.14 -444-3.5.J.2 ...J 3.5.J.3 g ( belefed) 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 12 EFPY 163 Part 1 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 14 EFPY 163a Part 2 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 16 EFPY 163b' Part 3 4.6-1 Chloride Stress Corrosion Test Results at 5007 164 6.1-1 (Deleted) .

6.2-1 (Deleted) -

1 I

Amendment No. 11,22,12,"i,72,71,"",0",100,*13,*1",1*7,101,137,1""li"2, 227, vii

.l .. .P 2.I  !!ASES (cout'd)

In order to ensure that the liel provided adequate pro- high local peaks, and because several rods must he Lection anals.:it the single rod withdrawat error, a moved to change power by a s18nificant percentage of rated pouer, tiie rate of power rise is very rasce of rni uithdtr.t.at accidentes w.as analyzed. Timis slow. Generally, the heat flux is in near equili-

.ana lysi s includcd start $stg the accitlent at various power In an assumed uniform Acvals. 'the naast severe sese involves an initial con ~

brines with the fission rate.

ditica in uhtch Line remeter is just subcritical and rod withdrawal approacia to the scrasi level, the - e the IMH syste.n is not yet se scale. This condition rate of po.scr riso is eso siore Elian 5 percent of e-(1ste. at quarter rod denetty. Additional conservatism rated power per aminute, and the APBH systein would ,

wcs L51:en in this analysis by assimiing that the IRet be enore clean edequate to assure a scrma before char.ne t clo se:st to the withdrawa rod is by-passed. the power could exceed the safety limit. The 15 i I

Thu results of this analysis show that the reactor is percent april scram reinains active inntti the mode scr.w.es d and peak pouer limited to one percent of switch is placed tai tlic RUN positiosi. 1his swit cla .

t ut e:I pouer, thus a aintalising HCPR above clie Safety occurs wht.n reactor pressure is greater than 850 psig. l Li:u t t . 1:ased un time above maalysis, the 1101 provides paotection against local control rod withdraual errors c. APRM Flux Scraun Trip Setting (Rune Mode _1 and continuous withdrawal of control rode in acquence an:i provides backup protection for the APitH. 1he APHH flux scran trip in the run mode consists of a flow referenced scrami serpoint and a fixed .

b. Alt!! *: lux Scrr.sn Trip Sett inct (Refinal or Startinp and . liigli neutron flux scrasa setpoint. The APRM f refercisced neutron flux signal is passed t ough a i

.;a. ;yg.,.._1gg )

filtering network witle a time constant ch is Foi operation in the startup anode while the reactor is representative of the fuel dyneinics. als provides. l at Ice pressure, the APITl scrasa setting of 15 percent a flow referenced signal tinit appr amates tiie l uf r..ted potser provides adequate clienral naargin between average heat flur. or thennal' pou that is developed i in the core during transit-nt steady-state cci.di- i t he ::etpoiat and tha safety 11a:lt, 25 percent of rated. f 1he c.amic is adequate to a:cc:staodato anticipated tions. This prevents spur a scramu, which have c: ant avers associated witti penser ple:it startup. Effects an adverse effect on re. or safety because of the i of fuercasing pressure at zero or low void content are resulting thenaal str - es. Exass ales of events init.:,r, cold t;cter f roia nources available during startup which can result in entary neutron fluc spikes l f c a st r.ucle coitier tisan thct already in the systein, are neonientary fl changes in the recirculation (teg ratere coef ficient s are sanall, and control rod systein flow, a s; mull press.ure distuibances '

pat terns are coautrained to be unifens by operating during turh e stop valve and turt.ine control i

3.racedures bashed up by the ro;l wortin minimizer and the valvo tes ag. Tisese fltet spikes represent no

• k.ul Seques:ce Cont rol Systein. Worth of inillvidual rods liaza rel o the fuel since they cre only of a fee In very f or in si ur. l fo rni rod pattern. Thus, of all sec s duration and less than 1207. of rated thernal pr:riuta teounces of reactivity input, unifor.a control p er.

r.v l ri

• irt r..v.i t is the mast prob.ille cause of signifi-e t.a p..rer rise. liccauae the flux distribution asno- The APatH , flow referenced scre:a trip se*

• u !

8-c l.ti .-d u tt h ual fosui i o.1 wit t.alrawals does not insolve recircielation s'l - .;3=tante up to 11R of Anendment No. ) ..aem

1 Insert a:

The APRM system obtains neutron flux input signals from LPRMs (fission chambers) and is j calibrated to indicate percent rated thermal power. The APRM scrams in the run mode are

)

! a flow referenced scram and a fixed high neutron flux scram. As power rises during '

transients, the instantaneous neutron flux (as a percentage of rated) will rise faster than the rate of heat transfer from the fuel (percentage of rated thermal power) due to the ,

thermal time constant of the fuel and core thermal power will be less than the power  !

indicated by the APRMs (neutron flux) at either scram setting.

? l The APRM flow referenced scram trip settmg, nominally varies from 54% power at 0%

recirculation flow to 120% power at 100% recirculation flow but is limited to 117% rated power. The flow referenced trip will result in a significantly earlier scram during slow j thermal transients, such as the loss of 80'F feedwater heating event, than would result l from the 120% fixed high neutron flux scram. The lower flow referenced scram setpoint therefore decreases the severity (ACPR) of a slow thermal transient and allows lower MCPR Operating Limits if such a transient is the limiting abnormal operational transient l during a certain exposure interval in the cycle. The flow referenced trip also provides I

protection for power oscillations which may result from reactor thermal hydraulic instability.

l The APRM fixed high neutron flux scram protects the reactor during fast power increase transients if credit is not taken for a direct (position) scram or flow referenced scram.

l l

l

JAFNPP 2.1 BASES (Cont'd) from the MCPR operating limits specified in the Core Operating Limits Report.

c. APRM Flux Scram Trip Setting (Run Mode) (cont"d) d. APRM Rod Block Trip Setting ratM pows. This & Row rdwW @ Reactor power level may be varied by moving control rods WM anhscrarn du% h MM or by varying the recirculahon flow rate. The APRM tr s,M asb W M hatw system provides a control rod block to prevent rod

, h 8x neutmn withdrawal beyond a gwen point at constant recirculation scrarn $ h h h h scrarn @ flow rate, and thus provides an added level of prnancem theh hennesh M ( dahM before APRM Scram. This rod block trip sethng, wtuch is trN W aNows h @ hMa automatically varied with recirculahon loop Sow rate trN h EmMng operm WW prevents an increase in the reactor power level to dunng a e exposure m cycle. excessive values due to control withdrawal. The now i

The APRM fixed Aux signaldoes not variable trip setting parallels that of the APRM Scram and incorporale the constant,but responds diracuy to provides margin to scram, assuming a steady-state i instantaneous on Aux. This scram seapoint scrams operation at the trip setting, over the entire recirenlahon the reactor fast power increase transients il credit is Sow range. The actual power distrh dm in the core is not a drect (poestion) scram, and also serves to estabhshed by specified control rod sequences and is scr the reactor il credit is not taken for the Sow monitored conhnuously by the in-core LPRM system. As r enced scram. with the APRM scram trip setting, the APRM rod t20ck trip mitting is a4usted downward if the maxunum fraction of The scram trip setting must be aqusted to ensure that the hmeting power density exM the fraction of rated power, M transient peak is not increased fw any combina60n thus preserving the APRM rod block margin. As with the d maximum iracdon of hmMng power dansky M saam seueng, this rnay be accomplished by aqushng the and reactor core thermal power. The scram setting is 98i"'

l aqusted as specified in Table 3.1-1 when the MFLPD is greater than the fraction of rated power (FRP). This 2. Reactor Water low Level Scram Trip Setting aqustment may be accomphshed by either educ6ng the APRM scram and rod block settings or usung The reactor low water level scram is set at a point which will the indicated APRM segnal to reRect the high peaking assure that the water level used in the Bases for the Safety Umit condihon. is maintained. The scram setpoint is based on normal operating tempwatwe Wesswe Mhs hse Mvd Analyses of the limiting transients show that no scram s My canpensatd adjustment is required to assure that the MCPR will be greater than the Safety Umit when the transient is initiated i

Amendment No. 3df, LAG -%2-)

18

l JAFNPP .

TABL E 3.1-1 i

REACTOR PROTECTION SYSTras (Sca^* INSTRa_wNTATION REQUIRERAENTS Mode in Which Function Muumum No. of Must be Operable Total Number of Operable instrument -

instrument Channels Channels Per Refuel Startup. Run Provided by Design Trip System for Both Trip Systems Action (Note 3)

Trip Function Trip Level Settang (Note 7)

(Notes 1 and 2)

X X 1 Mode Switch A X

1 Mode Switch in Shutdown X X 2 A X

1 Manual Scram X 8 A s 96 % (120/125) X t 3 IRM High Flux of fus scale X 8 A X

3 IRM Inoperative 6 A s 15% Power X X 2 APRM Neutron Flux- ,

Startup (Note 15)

X 6 A or B ,

i 2 APRM Flow Referenced (Note 12)

Neutron exceed 117%) Flux (Note (Not to/13 '

pd M)

X 6 A or B 2 APRM Fixed High s 120% Power Neutron Flux GMc, 4j '

X 6 A or B X X 2 APRM inoperative (Note 10) s Amendment No. /, /,1%,4;p., 40 a

i JAFNPP -

i TABLE 3.1-1 (cont'd)

RFACT Mt PROTECTKW SYSus sa reRm RNSTRa"""*4TATION REQtilRpaaENTS 4

7

12. The APRM Flow Referansed Neutron Flux Scram setting shall be less than or equal to the Smit specified in the Core Operating Limits Report. - 1

13. The Avera0e Power Range Monitor scram funcien is varied as a funcbon of recirculation flow (W). The trip setting of this funct .

must be mamtained as speclSed in the Core Operating Lets Report.

C'.' thed 4 C "x; ST M 'O. -_^ x $_ _ 'Zd b ":d tr$ = 5: :=tr! 909 Of re he'd; S^f=  :::=i"=:( TM A.

14. TM A. " '

' ^: i ^ :: c- zT^ M' : ;:2 6-:ft te h'rtr::= Mieled .

.W.  : ^ x $_ :'yd f::: ; ^ '  :: i:

15. This Average Power Range Monitor scram funcemn is fixed point and is increased when the reactor mode swF.ch is placed in ,

Run poseen. t i

16. Instrumentaban common to PCIS.

a r

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t Amendment No. IN. M i 43a

,o,

Jaspy 3.5 (coat'd) 4.5 (coat'd)

J. Thermal Hydraulic Stab (11tr J. Theresl Erdres11c Stab 111tr

1. Whenever the reactor is la the startup o 1. Establish baseline and LPRIt neutcom flux run modes, two Reactor Coolant Sys moise values wit 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of enterlag the recirculation loops shall be la operat , region for ch moottoring is required with: maless b intag has been performed since the refueltag outage. Detector levels &

a. Total core flow greater than or a C of one Lytti string per core octant plus equal to 45 percent of cated, or detectors A and C of one LPRH ste*ng la the center of the core should be monitored.

b. Thermal power less than oc equal to the limit speelfi la Figure 3.5-1 (Line A).

except as specified a specifications 3.5.J.2 and 3.5.J.3.

2. With tuo Reactor sat System recircula-tion Icope in opor los and total core flow less than 45 per c,at of cated, and thermal power greater as the limit speelfied la Figure 3.5-1 ine A); or with one Reactor Coolant syst loop . operettas and thermal power gree e than the limit specifled la Figure 3. 1 (Line A):

a. Det ime the APSIE and LpBE aoise levels:

. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after reachlag steady- ,

state within the regions of Figure 3.5-1 where maaltoring is required, and at least once per S hours a thereafter; and Insed b .

mammt . p.#,

124a

JAFNPP 3.5 (cont *d)

2. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after completing on increase '

thermal power of 5 percent or more of ret thermal power.

b. If the APRM and LPRM neutron flux noi s are greater than 5 percent and poeter se times their established baseline noise ' inte corrective action within 15 minut o restore the noise levels to within the s within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, by lis:::'3, core flow /or reducing thermal power.

3. If during single-loop opere core thermal power is greater then the limit de by line A of Figure 3.5-1, and core flow is less 9 percent,immediately initiate corrective to restore core thermal power and/or core flow to the limits, specified in Figure 3.5-1, by incree core flow and/or initioting on ordorfy reduction of cor thermal power by inserting control rods.

4. The requir a applicable to single-loop operation in Specifica ' 1.1.A 2.1.A 3.1.A, 3.1.8, 3.2.C and 3.5.H be in effect within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the rem of one recirculation loop from service, or the re or shell be placed in at loest the hot shutdown ,

ition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I Amendment No. g,192 124b

i JAFNPP .  :

3.5 (cont'd) I resumption of two-loop ation

5. Durlas operation, folloulas a period of slagle-1 the discharge valve of low-speed pump l shall act be opened es the speed of the faster m to than 50 percent of it's  ;

rated space ao Reactor Coolant System Recircalation j

6. W '

oop la service, the reactor shall he pieced ta Not Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. i l

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Amendment No. 98, 124c j

o i

- . . - . . - . . . . . _ _ - -. . _ _ _ , _ = _ . .

Insert b:

J. Thermal Hvdraulic Stability

1. When the reactor is in the run mode:

a. Under normal operating conditions the reactor shall not be intentionally operated within the Power / Flow Exclusion Region defined in the Core Operating Limits Report (COLR).

b. If the reactor has entered the Power / Flow Exclusion Region, the operator shall immediately insert control rods and/or increase recirculation flow to establish operation outside the region.

K. Sinale-Loon Ooeration

1. The reactor may be started and operated, or reactor operation may continue, with a single Reactor Coolant System recirculation loop in operation. The requirements applicable to single-loop operation in Specifications 1.1.A, 2.1.A,3.1.A, 3.1.B, 3.2.C, and 3.5.H shall be in effect within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or the reactor shall be placed in at least the hot shutdown mode within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. During resumption of two-loop operation following a period of single-loop operation, the discharge valve of the lower speed pump shall not be opened unless the speed of the faster pump is less than 50 percent of its rated speed.

3. With no Reactor Coolant System recirculation loop in service, the reactor shall be placed in at least the hot shutdown mode within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

. .. - .. -.- . ~ - . - ~ - - - - - - ~ . . . . - .

9 s i I

JApNpp ,

3.5 BAsga (coat'd) ,

J. Thermal Evdraulle Stab 111tr operation in certata regions of the power v gh '

flow curve have been identified as having a potential for thesmal hydraulic test lity i

(Figure 3.5-1). These regions are loca la the high power / low flew area of the curv and can be encountered durlag startup, utdown, rod seguence exchange or reeless los pump trip.  ;

e associated with Operation la these regless flus moise levels. (- L g h higher than normal nest Increased awareness of and ApWE signal noise  :

whom operating la se regions will identify lastability and ow operator action to correct the problem. The neutros fluz moise level, thermal and core flow limits are prescribed la acco ce with the reeemosadations of General Elect Servlee Informaties Letter No. 380, Rev los 1 "aut Core Thermal Eydraulio  !

ability", dated February 10, 1984.

4- k. 61 Loco ope < cAcn t Requirlag the discharge valve of the lower speed i loop to remala closed nacil the speed of the faster pump is below 50 percent of its rated  !

speed provides assurance when solag from one to -

f two pump operation that excesolve vibration of the jet pump risers will not occur. i w .e,er...es

1. "Fitapatrick Nuclear power plant Single-Loop I Operation". NEDO-24281. August 1940. i 1 icc. on ok kw ic.~, c. E xclu ices s S O w - %

,,, a s<a sus sm +

w u wn n. nkpcAc;ck u tect L e< Plc C .

m i

(,E.M- W -OM-0735, pghgug gg$

AnsadmentNo.[. .M. 131 i

a ,

c -

Insert c:

10 CFR 50, Appendix A, General Design Criterion 12 requires that power oscillations are either prevented or can be readily detected and suppressed without exceeding specified fuel design limits. To minimize the likelihood of a thermal hydraulic instability which results in power oscillations, a power / flow exclusion region to be avoided during normal operation is calculated using the approved methodology specified in Technical Specification 6.9(A)4. Since the exclusion region may change each fuel cycle, the limits are contained in the Core Operating Limits Report. Specific directions are provided to avoid operation in the exclusion region and to immediately exit the region if entered.

Entries into the exclusion region are not part of normal operation, but may result from an abnormal event, such as a single recirculation pump trip or loss of feedwater heating, or be required to prevent equipment damage. In these events, time spent within the exclusion region is minimized.

Although operator actions can prevent the occurrence of and protect the reactor from an instability, the APRM flow-biased reactor scram will suppress power oscillations prior to exceeding the fuel safety limit (MCPR). Reference 3.5.L.2 demonstrated that this protection is provided at a high statistical confidence level for core-wide mode oscillations and at a nominal statistical confidence level for regional mode oscillations. This reference also demonstrated that the core wide mode of oscillation is preferred due to the large single-phase channel pressure drop associated with the small fuel inlet orifice diameters.  ;

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Figure 3.5-1

• Thermal Power and Core Flow Listits of Specifications 3.5.J.1, 3.5.J.2 and 3.5.J.3  :

70  ;

Stability Stability Monitoring [

tenitoring (APRM and LPRM) Requ ed '

Stability Monitoring (APitt and For Single Loop l (APHIt and LPpM) Required Ipfel) Operation 60 _ During Two-Loop Operation Required  ;

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During Line A I e Single and  ;

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50 Operah Singla--Loop Operation j U Prohibited I f

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• ii iiiiiii i i 70 40 45 50 60 30 CORE FLOW (PERCENT RATED)

Amendment No. [ ,d J ,JPI, [ ,

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- .. = - - . - - . - - . - . . ---- - - - - .-. ._ __

..,., JAFNPP (A) ROUTINE REPORTS (Continued)

4. CORE OPERATING UMITS REPORT

a. Core operating limits shall be established prior to startup from each reload cycle, or prior to any remaining portion of a reload cycle for the following:

. The Average Planar Unear Heat Generation Rates (APLHGR) of Specification 3.5.H;

. The Minimum Critical Power Ratio (MCPR) and MCPR low flow adjustment factor, K,, of Specifications 3.1.8 and 4.1.E;

. The Unear Heat Generation Rate (LHGR) of Specification 3.5.l;

. The Reactor Protection System (RPS) APRM flow biased trip settings of Table 3.1 1;4md-

. The flow biased APRM and Rod Block Monitor (RBM) rod block setti l

andmshall

. T ofbehrw TableE3.2

/ 3 ed c.,n J QecSc< L 3.5.3..

j umented xcb in $1euCore 42es: Operating Umits Report (COLR) l

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC as described in:

1. " General Electric Standard Application for Reactor Fuel," NEDE-24011 P, latest approved version and amendments.

2. " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR LCCA Loss of-Coolant Accident Analysis," NEDC-31317P, October,1986 including latest errata and addenda.

3. " Loss of-Coolant Accident Analysis for James A. FitzPatrick Nuclear Power Plant," NEDO 21662 2, July,1977 including latest errata and addenda.

A

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanecal limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident l analysis limits) of the safety analysis are met. 7 wek

d. The COLR, including any mid-cycle revisions or supplements thereto, shall 4 be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident S li=-;+:4 .

m.uem ,<s_s w u-ve ,m e i E00 - 3n S60- A , he 1991

5. " BLR owned G<cq Lcg- b u Shkkk bM("5 LM*M*3 I

OcMoest ," rE Do - 3596o- 4, Sq eu<M l i , Mc.< ck 1992 Amendment No. 4.Q.

254d

JAFNPP l .

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Amendment No. 32,170 e 2544 * ,caJ^" i$ e , S 1