ML20134E967

Amend 236 to License DPR-59,revising TS to Establish Operability Requirements for Avoidance & Protection from Thermal Hydraulic Instabilities to Be Consistent W/Bwrog long-term Solution Option I-D

ML20134E967
Person / Time
Site:	FitzPatrick
Issue date:	10/30/1996
From:	Bajwa S NRC (Affiliation Not Assigned)
To:
Shared Package
ML20134E970	List: ML20134E967 ML20134E971
References
NUDOCS 9611040203
Download: ML20134E967 (15)

Text

. - _ _.

- - ~ ~ _

'p 2cer'

\\

UNITED STATES y

g j

NUCLEAR REGULATORY COMMIS810N WASHINGTON, D.C. enmaa m k...../

POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK WCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.236 License No. DPR-59 1.

The Nuclear Regulatory Commission (the Comission) has found that:

A.

The application for amendment by Power Authority of the State of New York (the licensee) dated March 22, 1996, as supplemented October 11, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized L

by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comen defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

l 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating Licensa No. DPR-59 is hereby amended to read as follows:

i 9611040203 961030 j

PDR ADOCK 05000333 l

P PDR

4 i -

(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.236, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance to be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION f

S. Singh Bajwa, Acting Director Project Directorate I-I Division of Reactor Projects - I/II i

Office of Nuclear Reactor Regulation Attachment.

Changes to the Technical Specifications Date of Issuance: October 30, 1996 i

l l

i l

ATTACHMENT TO LICENSE AMENDMENT NO.236 I

l FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Revise Appendix A as follows:

Remove Paaes Insert Paoes 11 11 vii vit 17 17 18 18 40 40 43a 43a 124a 124a 124b Delete 124c Delete 131 131 134 134 254c 254c 254d 254d 254e 254e,f i

254f 254e,f i

1 I

4 I

I

JAFNPP l

l 4

TABLE OF CONTENTS (Cont'd) i EASA F.

ECCS-Cold Condition F.

122 G.

Maintenance of Filled Discharge Pipe G.

1.22a i-H.

Average Planar Linear Heat Generation Rate H.

123 4

(APLHGR) i I.

Linear Heat Generation Rate (LHGR) 1.

124 J.

Thermal Hydraulic Stability DELETED 124a K.

Single-Loop Operation NONE 124a l

SURVElLLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 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 Safety / Relief Valves E.

142a F.

Structural integrity F.

144 j

G.

Jet Pumps G.

144 H.

DELETED l

t.

Shock Suppressors (Snubbers) 1.

145b i

3.7 Containment Systems 4.7 165 A.

Primary Containment A.

165 B.

Standby Gas Treatment System B.

181 C.

Secondary Containment C.

184 D.

Primary Containment isolation Valves D.

185 3.8 Miscellaneous Radioactive Material Sources 4.8 214 3.9 Auxiliary Electrical Systems 4.9 215 A.

Normal and Reserve AC Power Systems A.

215 B.

Emergency AC Power System B.

216 C.

Diesel Fuel C.

218 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 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.

DELETED C.

Battery Room Ventilation C.

239 Amendment No. 2C,12,90,112,1SS, 231, 236 ii

- ~. _ - -

JAFNPP LIST OF FIGURES t

Fiaures Ilt!g Eggg 4.1-1 (Deleted) 4.2-1 (Deleted) 3.4-1 Sodium Pentaborate Solution (Minimum 34.7 B-10 Atom % Enriched) 110 Volt:me-Concentration Requirements i

l 3.4-2 Saturation Temperature of Enriched Sodium Pentaborate Solution 111 3.5-1 (Deleted) 3.6 's Reactor Vessel Pressure - Temperature Limits Through 12 EFPY 163 4

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 j

Part 3 4.6 1 Chloride Stress Corrosion Test Results at 500'F 164 i

6.1-1 (Deleted) 6.2-1 (Deleted) f f

f i

a

+

i i

Amendment No. ' i, 22,12, Si, 72, 'i, SS, 00,109, ' 13,

• 1 S,1 * ?,131,137,1 SS,15 2, 227,236 vii

l l

l JAFNPP i

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 i

protection against the single rod withdrawal error, a range assumed uniform rod withdrawal approach to the scram l

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 i

most severe case involves an initial condition in which the more than adequate to assure a scram before the power i

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 i

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 indicata continuous withdrawal of control rods in sequence and percent rated thermal power. The APRM scrams in the run i

provides backup protection for the APRM.

mode are a flow referenced scram and a fixed high neutron l

flux scram. As power rises dunng transients, the b.

APRM Flux Scram Trio Settina (Refuel or Startuo and Hot instantaneous neutron flux (as a percentage of rated) will Standby Mode) rise faster than the rate of host 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.

l 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 patterns are constrained to be such as the loss of 80*F feedwater heating event, than i

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 F

Amendment No. 43, 236 17

_m.-

1 i

JAFNPP 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 j

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 l

hydraulic instability.

the reactor power level to excessive values due to I

control withdrawal. The flow variable trip setting 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 rd 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 l

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, f

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 Settino 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 level 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 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 i

Reactor power level may be varied by moving control l

rods or by varying the recirculation flow rate. The Amendment No. 49, ' ? 9, 'S2, 236 18 l

1

. ~.

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 Refuel Startup Run Provided by Design Trip System (Notes 1 and 2)

Trip Function Trip Level Setting (Note 7) for Both Trip Systems Action (Note 3) 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 exceed 117%) (Note 13) 2 APRM Fixed High 5120% Power X

6 A or B Neutron Flux 2

APRM inoperative (Note 10)

X X

X 6

A or B Amendment No. , 18,183, 227, 236 i

40

JAFNPP TABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAMI INSTRUMENTATION REQUIREMENTS

12. The APRM Flow Referenced Neutron Flux Scram setting shall be less than or equa! 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.

14. Deleted.

i

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.

t t

i i

l i

1 l

i i

Amendment No. 103, 227,236 l

43a l

JAFNPP 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 regior.

K.

Sinole-Loon Operation 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 l

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, i

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. 39788, 236 124a

JAFNPP 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 renion to be avoided during normal operation is calculated using the approved L.

References methodology spec 6Hed in Technical Specification 6.9(A)4.

Since the exclusiv.. region may chF,gu each fuel cycle, the 1.

"FitzPatrick Nuclear Power Plant Single-Loop limits are contained in the Core Omang Limits Report.

Operation", NEDO-24281, August 1980.

Specific directions are provided to avoid operation in the exclusion region and to immediately exit the repim if 2.

" Application of the ' Regional Exclusion with Flow-entered. Entries into the exclusion region are not rart of Biased APRM Neutron Flux Scram' Stability Solution normal operation, but may result from an abaor.T.y event.

(Option I-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 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.

Amendment No. " Si, 98, 236 131

JAFNPP (THIS PAGE INTENTIONALLY BLANK)

Amendment No. ? ', 20, 52, s', 98, 236 134

i j

JAFNPP i

i (A)

ROUTINE REPORTS (Continued) 4.

CORE OPERATING LIMITS 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 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.B and 4.1.E; The Linear Heat Generation Rate (LHGR) of Specification 3.5.l; The Reactor Protection System (RPS) APRM flow biased trip settings of Table 3.1-1; I

j The flow biased APRM and Rod Block Monitor (RBM) rod block settings of Table 3.2-3; and l

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

I and shall be documented in the Core Operating Limits Report (COLR).

i

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 - LOCA 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.

4. "BWR Gwners' 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.

l Amendment No. 443, 236 254c

JAFNPP I

l

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

I

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.

Amendment No. 32,'10,102, 236 254d I

i r

JAFNPP l

i l

l t

(THESE PAGES INTENTIONALLY BLANK) 1 4

l 1

i 1

i d

I l

l l

1 Amendment No. 32,110,1S2, 236 254e,f l

l l