ML20132C333
| ML20132C333 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/06/1996 |
| From: | Miraglia F NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20132C337 | List: |
| References | |
| NUDOCS 9612180288 | |
| Download: ML20132C333 (38) | |
Text
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i uo vy UNITED STATES l
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NUCLEAR REGULATORY COMMISSION WASHINoTON, D.C. 20065-0001 i
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POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET N0. 50-333 i
l JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 239 License No. DPR-59 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Power Authority of the State of New York (the licensee) dated June 12, 1992, as supplemented September 17, 1992, March 17, 1993, August 17, 1993, August 18, 1993, December 29, 1993, and June 29, 1995, and August 15, 1996, October 3, 1996, October 23, 1996, November 14, 1996, November 20, 1996 (JPN 1 045), November 20, 1996 (JPN-96-046), and November 27, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission'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 Commission; i
C.
There is reasonable assurance (1) that the activities authorized 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 Commission's regulations; I
l D.
The issuance of this amendment will not be inimical to the common l
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 Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(1), 2.C.(2), 2.E., 2.E.(1), 2.E.(2), and 2.E(3) of Facility Operating License No. DPR-59 are hereby amended to read as i
follows:
9612180288 961206 PDR ADOCK 05000333 P
l i l
l 3.
This license amendment is effective as of the date of its issuance to be implemented upon plant startup following the refueling outage for cycle 13.
FOR THE NUCLEAR REGULATORY COMMISSION
$/)
f
- n J. M a ia, Acting Director
/
f ce of lear Reactor Regulation Attachments:
1.
Pages 3 and of License 2.
Changes to the Technical Specifications Date of Issuance:
December 6, 1996 1
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\\
1 i
i i
l ATTACHMENT 1 TO LICENSE AMENDMENT N0.239 1
FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 1
2 Revise the license as follows:
Remove Pace Insert Paae 3
3 4
4 b
1 l
~.
~ _
I i (3)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use,.at any time, any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct, source and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus, components or tools.
(5)
Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
The license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59,of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules,.
regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2536 megawatts (thermal).
l (2)
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.4, are hereby incorporated in the license.
The licensee shall l
operate the facility in accordance with the Technical Specifications.
l
- Each amendment updates this paragraph to indicate the latest amendment to the License.
l
. D.
Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:
"FitzPatrick Modified Amended Security Plan," with revisions submitted through March 7, 1988; "FitzPatrick Modified Amended Security Force Training and Qualification Plan," with revisions submitted through April 10, 1985; and "FitzPatrick Security Contingency Plan," with revisions submitted through June 20, 1980. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
E.
Power Vorate License Amendment Imolementation The licensee shall complete the following actions as a condition of the approval of the power uprate license amendment.
(1) Recirculation Pumo Motor Vibration Perform monitoring of recirculation pump motor vibration during initial Cycle 13 power ascension for uprated power conditions.
(2) Startuo Test Proaram The licensee will follow a startup testing program, during Cycle 13 power ascension, as described in GE Licensing Topical Report NEDC-31897-P-1, " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate." The startup test program includes system testing of such process control systems as the feedwater flow and main steam pressu+re control systems. The licensee will collect steady-state operational data during various portions of the power ascension to the high u licensed power level so that predicted equipment performance characteristics can be verified. The licensee will do the startup testing program in accordance with its procedures.
The licensee's approach is in conformance with the test guidelines of GE Licensing Topical Report NEDC-31897P-1, " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," June 1991 (proprietary), GE Licensing Topical Report NE00-31897, " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," February 1992 (nonproprietary), and NEDC-31897P-A, Class III (proprietary), May 1992.
(3) Human Factors The licensee will review the results of the Cycle 13 startup test program to determine any potential effects on operator training.
Training issues identified will be incorporated in Licensed Operator training during 1997. Simulator discrepancies identified will be addressed in accordance with simulator Configuration Management procedural requirements.
ATTACHMENT 2 TO LICENSE AMENDMENT NO. 239 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Revise Appendix A as follows:
l I
Remove Paaes Insert Paaes 2
2 5
5 l
6a 6a 11 11 14 14 l
15 15 l
16 16 i
19 19 l
20 20 l
27 27 l
29 29 I
34 34 35 35 41 41 43 43 100 100 102 102 117 117 121a 121a 125 125 139 139 142a 142a 147 147 149 149 152 152 166 166 187 187 188 188 l
188a 188a 193 193 254c 254c 285 285 285a 285a
i JAFNPP 1.0 (cont'd)
C.
Cold Condition - Reactor coolant temperature.5212*F.
4.
Instrument Check - An instrument check is a qualitative determination of acceptable operability F
D.
Hot Standby Condition - Hot Standby condition means by observation of instrument behavior during operation with coolant temperature > 212'F, the Mode operation. This determination shall include, where Switch in Start-up/ Hot Standby and reactor pressure possible, comparison of the instrument with other l
< 1,040 psig.
independent instruments measuring the same variable.
E.
Immediale - Immediate means that the required action will be initiated as soon as practicable considering the safe 5.
Instrument Channel Functional Test - An instrument operation of the unit and the importance of the required channel functional test means the injection of a action.
simulated signal into the instrument primary sensor where possible to verify the proper instrument F.
Instrumentation channel response, alarm and/or initiating action.
1.
Functional Test - A functional test is the manual 6.
Primary Containment isolation Actuation
)
operation or initiation of a system, subsystem, or Instrumentation Response Time for Main Steam Line component to verify that it functions within design isolation is the time interval which begins when the tolerances (e.g., the manual start of a core spray monitored parameter exceeds the isolation actuation pump to verify that it runs and that it pumps the set point at the channel sensor and ends when the required volume of water).
Main Steam isolation Valve solenoids are de-energized (16A-K14, K16, K51, & K52 pilot 2.
Instrument Channel Calibration - An instrument solenoid relay contacts open). The response time channel calibration means the adjustment of an may be measured in one continuous step or in instrument signal output so that it corresponds, overlapping segments, with verification that all within acceptable range, and accuracy, to a known components are tested.
value(s) of the parameter which the instrument i
monitors. Calibration shall encompass the entire 7.
Logic System Function Test - A logic system instrument channel including actuation, alarm or functional test means a test of relays and contacts trip.
of a logic circuit from sensor to activated device to ensure components are operable per drisign intent.
3.
Instrument Channel - An instrument channel means Where practicable, action will go to completion: i.e.,
an arrangement of a sensor and auxiliary equipment pumps will be started and valves ope'tated.
required to generate and transmit to a trip system a l
single trip signal related to the plant parameter 8.
Protective Action - An action initiated by the i
monitored by that instrument channel.
Protection System when limiting safety system setting is reached. A protective action can be at a channel or system level.
Amendment No. 3, 134, ?83, 239 2
i
JAFNPP 1.0 (cont'd) opened to perform necessary operational activities.
deficiency subject to regulatory review.
2.
At least one door in each airlock is closed and S.
Secondary Containment Intearity - Secondary containment sealed.
integrity means that the reactor building is intact and the 3.
All automatic containment isolation valves are operable or de-activated in the isolated position.
1.
At least one door in each access opening is closed.
4.
All blind flanges and manways are closed.
2.
The Standby Gas Treatment System is operable.
I N.
Rated Power - Rated power refers to operation at a reactor 3.
All automatic ventilation system isolation valves are I
power of 2,536 MWt. This is also termed 100 percent operable or secured in the isolated position.
i power and is the maximum power level authorized by the operating license. Rated steam flow, rated coolant flow, T.
Surveillance Freauency Notations / Intervals l
rated nuclear system pressure, refer to the values of these i
parameters when the reactor is at rated power (Reference The surveillance frequency notations / intervals used in these 1).
specifications are defined as follows:
O.
Reactor Power Operation - Reactor power operation is any Notations Intervals Freauency operation with the Mode Switch in the Startup/ Hot i
Standby or Run position with the reactor critical and above D
Daily At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1 percent rated thermal power.
W Weekly At least once per 7 days M
Monthly At least once per 31 days P.
Reactor Vessel Pressure - Unless otherwise indicated, O
Quarterly or At least once per 92 days reactor vessel pressures listed in the Technical every 3 months Specifications are those measured by the reactor vessel SA Semiannually or At least once per 184 days i
steam space sensor.
every 6 months A
Annually or Yearly At least once per 366 days Q.
Refuelina Outaae - Refueling outage is the period of time 18M 18 Months At least once per 18 months (550 between the shutdown of the unit prior to refueling and days) the startup of the Plant subsequent to that refueling.
R Operating Cycle At least once per 24 months (731 days)
R.
Safety Limits - The safety limits are limits within which S/U Prior to each reactor startup the reasonable maintenance of the fuel cladding integrity NA Nat applicable and the reactor coolant system integrity are assured.
Violation of such a limit is cause for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it i
indicates an operational Amendment No. M, 134,188, 227, 233, 239 5
Core Operatina Limits Report (COLR) l Z.
Too of Active Fuel This report is the plant-specific document that The Top of Active Fuel, corresponding to the top of provides the core operating limits for the current i
the enriched fuel column of each fuel bundle, is operating cycle. These cycle-specific operating i
located 352.5 inches above vessel zero, which is limits shall be determined for each reload cycle in the lowest point in the inside bottom of the reactor accordance with Specification 6.9.A.4. Plant vessel. (See General Electric drawing No.
operation within these operating limits is addressed 919D690BD.)
in individual Technical Specifications.
i AA.
Rod Density AE.
References Rod density is the number of control rod notches 1.
General Electric Report NEDC-32016P-1, inserted expressed as a fraction of the total number
" Power Uprate Safety Analysis for James A.
of control rod notches. All rods fully inserted is a FitzPatrick Nuclear Power Plant," April 1993 7
condition representing 100 percent rod density.
(proprietary), including Errata and Addenda Sheet No.1, dated January 1994.
l AB.
Purae-Puraina Purge or Purging is the controlled process of discharging air or gas from a confinement in such a manner that replacement air or gas is required to l
purify the confinement.
AC.
Ventina Venting is the controlled process of releasing air or gas from a confinement in such a manner that replacement air or gas is not provided or required.
v Amendment No. 75, 93,1S2, 239 6a
JAFNPP 2.1 (cont'd) 2.
Reactor Water Low Level Scram Trio Settina Reactor low water level scrim setting shall be 2177 in. above the top of the active fuel (TAF) at normal operating conditions.
3.
Turbine Stoo Valve Closure Scram Trio Settina Turbine stop valve scram shall be s10 percent valve closure from full open when the reactor is at or above 29% of rated power.
4.
Turbine Control Valve Fast Closure Scram Trio Settina Turbine control valve fast closure scram control cil pressure shall be set at 500 <P <850 psig.
5.
Main Steam Line Isolation Valve Closure Scram Trio Settina Main steam line isolation valve closure scram shall be s10 percent valve closure from full open.
6.
Main Steam Line Isolation Valve Closure on Low Pressure t
When in the run mode main steam line low pressure initiation of main steam line isolation valve closure shall be 2825 psig.
l i
i Amendment No. 27, 29,'19, 239 11
JAFNPP 1.1 BASES (Cont'd) l E.
References C.
Power Transient 1.
" General Electric Standard Application for Plant safety analyses have shown that the scrams Reactor Fuel," NEDE-24011-P-A-13, August caused by exceeding any safety system setting will 1996.
assure that the Safety Limit of 1.1.A or 1.1.B will not be exceeded. Scram times are checked periodically to 2.
FitzPatrick Nuclear Power Plant Single-Loop assure the insertion times are adequate. The thermal Operation, NEDO 24281, August 1980.
power transient resulting when a scram is t
accomplished other than by the expected scram signal 3.
GE12 Compliance with Amendment 22 of (e.g., scram from neutron flux following closure of the NEDE-24011-P-A (GESTAR 11), NEDE-32417P, main turbine stop valves) does not necessarily cause December 1994.
I fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only i
accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is i
supported by the extensive plant safety analysis.
D.
Reactor Water Level (Hot or Cold Shutdown Condition)
During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds the core height. Establishment of the Safety Limit at 18 in. above the top of the fuel provides adequate margin. This level will be continuously monitored whenever the recirculation t
pumps are not operating.
{
Amendment No. 'i, 98,1S2. 238, 239
JAFNPP BASES 2.1 FUEL CLADDING INTEGRITY The most limiting transients have been analyzed to The abnormal operational transients applicable to determine which result in the largest reduction in operation of the FitzPatrick Unit have been analyzed CRITICAL POWER RATIO. The type of transients
' throughout the spectrum of planned operating evaluated were increase in pressure and power, l
conditions up to the thermal power condition of positive reactivity insertion, and coolant temperature a
i 2,536 MWt. The analyses were based upon plant decrease. The limiting transient yields the largest operation in accordance with the operating map given delta MCPR. When added to the Safety Limit, the in the current load line limit analysis. In addition, required operating limit MCPR in the Core Operating i
2,536 MWt is the licensed maximum power level of Limits Report is obtained.
FitzPatrick, and this represents the maximum steady-state power which shall not knowingly be exceeded.
The evaluation of a given transient begins with the system initial parameters shown in the current reload The transient analyses performed for each reload are analysis and Reference 2 that are input to the core I
described in Reference 2. Models and model dynamic behavior transient computer programs
(
conservatism are also described in this reference. As described in Reference 2. The output of these discussed in Reference 4, the core wide transient programs along with the initial MCPR form the input l
analysis for one recirculation pump operation is for the further analyses of the the'mally limited I
conservatively bounded by two-loop operation bundle with a single channel transient thermal analysis, and the flow-dependent rod block and scram hydraulic code. The principal result of the evaluation setpoint equations are adjusted for one-pump is the reduction in MCPR caused by the transient.
l operation. Reference 1 evaluates the safety significance of uprated power operation at 2,536 MWt. This evaluation is consistent with and
[
demonstrates the acceptability of the transient analyses required by Reference 2.
Fuel cladding integrity is assured by the applicable operating limit MCPR for steady state conditions j
given in the Core Operating Limits Report (COLR).
These operating limit MCPR's are derived from the established fuel cladding integrity Safety Limit, and an f
analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient.
j Amendment No. 19, Si, 72, 98,1 S2, 239 l
15 i
I
JAFNPP 2.1 BASES (Cont'd)
A.
Trio Settinos f
The MCPR operating limits in the COLR are conservatively assumed The bases for individual trip settings are discussed in the to exist prior to initiation of the transients.
following paragraphs.
This choice of using conservative values of controlling parameters 1.
Neutron Flux Trio Settinas i
cnd initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of a.
IRM Flux Scram Trio Settina control parameters and analyzing at higher power levels.
The IRM system consists of 8 chambers,4 in Steady-state operation without forced recircula' ion is not permitted.
each of the reactor protection system logic The analysis to support operation at various power and flow channels. The IRM is a 5-decade instrument rclationships has considered operation with either one or two which covers the range of power level recirculation pumps.
between that covered by the SRM and the APRM. The 5 decades are covered by the IRM i
in summary:
by means of a range switch and the 5 decades are broken down into 10 ranges, each being The abnormal operational transients were analyzed to one-half of a decade in size. The IRM scram the licensed maximum power level.
trip setting of 120 divisions is active in each range of the IRM. For example, if the l
The licensed maximum power level is 2,536 MWt.
instrument were on Range 1, the scram setting would be a 120 divisions for that range; Analyses of transients employ adequately likewise, if the instrument were on range 5, i
conservative values of the controlling reactor the scram would be 120 divisions on that parameters.
range. Thus, as the IRM is ranged up to accommodate the increase in power level, the i
The analytical procedures now used result in a more scram trip setting is also ranged up. The most logical answer than the alternative method of significant sources of reactivity change during assuming a higher starting power in conjunction with the power increase are due to control rod the expected values for the parameters.
withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods, that heat flux is in equilibrium with the neutron flux and an IRM scram would result in a reactor shutdown well l
before any Safety Limit is exceeded.
I i
i Amendment No. li, 18, 21, 30,162,239 16 j
JAFNPP 2.1 BASES (Cont'd) 3.
Turbine Ston Valve Closure Scram Trio Settings 5.
Main Steam Line Isolation Valve Closure Scram Trio Settina The turbine stop valve closure scram trip anticipates the The low pressure isolation of the main steam lines at 825 pressure, neutron flux and heat flux increase that could psig was provided to give protection against rapid reactor result from rapid closure of the turbine stop valves. With a depressurization and the resulting rapid cooldown of the scram trip setting of s10 percent of valve closure from full vessel. Advantage was taken of the scram feature which open, the resultant increase in surface heat flux is limited occurs when the main steam line isolation valves are closed, such that MCPR remains along the Safety Limit even during to provide for reactor shutdown so that high power operation the worst case transient that assumes the turbine bypass is at low reactor pressure does not occur, thus providing closed. This scram is bypassed when reactor power is protection for the fuel cladding integrity safety limit.
below 29% of rated, as measured by turbine first stage Operation of the reactor at pressures lower than 825 psig i
pressure, consistent with the safety analysis discussed in requires that the Reactor Mode Switch be in the Startup Reference 1.
position where protection of the fuel cladding integrity safety limit is provided by the APRM high neutron flux scram and 4.
Turbine Control Valve Fast Closure Scram Trio Settina the IRM. Thus, the combination of main stream line low pressure isolation and isolation valve closure scram assures This turbine control valve fast closure scram anticipates the the availability of neutron flux scram protection over the pressure, neutron flux, and heat flux increase that could entire range of applicability of the fuel cladding integrity result from fast c osure of the turbine control valves due to safety limit. In addition, the isolation valve closure scram load rejection exceeding the capability of the turbine bypass.
anticipates the pressure and flux transients which occur The Reactor Protection System initiates a scram when fast during normal or inadvertent isolation valve closure. With closure of the control valves is initiated by the fast acting the scrams set at s10 percent valve closure, there is no solenoid valves. This is achieved by the action of the fast increase in neutron flux.
acting solenoid valves in rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc 6.
Main Steam Line isolation Valve Closure on Low Fressure dump valves. This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice The low pressure isolation minimum limit at 825 psig was logic input to the reactor protection system. This trip provided to give protection against fast reactor setting, a nominally 50 percent greater closure time and a depressurization and the resulting rapid cooldown of the different valve characteristic from that of the turbine stop vessel. Advantage was taken of the scram feature which valve, combine to produce transients very similar and no occurs when the main steam line isolation valves are closed more severe than for the stop valva. No significant change to provide for reactor shutdown so that operation at in MCPR occurs. Relevant transient analyses are discussed pressures lower than those specified in the thermal hydraulic in Section 14.5 of the Final Safety Analysis Report and safety limit does not occur, although operation at a pressure Reference 1. This scram is bypassed when reactor power is lower than 825 psig would not necessarily constitute ari below 29 percent of rated, as measured by turbine first unsafe condition.
stage pressure.
Amendment No. 'i, 21, 30, 37, 43. 239 19
=
JAFNPP 2.1 BASES (Cont'd)
B.
Not Used C.
References 1.
General Electric Report, NEDC-32016P-1, " Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant," April 1993 (proprietary), including Errata and Addenda Sheet No.1, dated January 1994.
2.
" General Electric Standard Application for Reactor Fuel,"
NEDE-24011-P-A-13, August 1996.
3.
(Deleted) 4.
FitzPatrick Nuclear Power Plant Single-Loop Operation, NEDO-24281, August,1980.
Amendment No. ,19, Si, SS,162,190, 239 20 (Next page is 23)
~
t k
JAFNPP 1.2 REACTOR COOLANT SYSTEM 2.2 REACTOR COOLANT SYSTEM APPLICABILITY:
APPLICABILITY:
Applies to limits on reactor coolant system pressure.
Applies to trip settings of the instruments and devices which are provided to prevent the reactor coolant system safety limits from being exceeded.
OBJECTIVE:
OBJECTIVE:
To establish a limit below which the integrity of the Reactor Coolant To define the level of the process variables at which automatic System is not threatened due to an overpressure condition.
protective action is initiated to prevent the safety limits from being exceeded.
SPECIFICATION:
SPECIFICATION:
1.
The reactor vessel dome pressure shall not exceed 1,325 psig 1.
The Limiting Safety System setting shall be specified below:
j at any time when irradiated fuel is present in the reactor i
vessel.
A.
Reactor coolant high pressure scram shall be _< 1,080 l
psig.
B.
At least 9 of the 11 reactor coolant system safety / relief valves shall have a nominal setting of 1,145 psig with l
an allowable setpoint error of $_3 percent.
i
?
Amendment No. 1 S, 30,15, Si, S9, 217, 239 27
i JAFNPP 1.2 and 2.2 BASES i
The reactor coolant pressure boundary integrity is an important The limiting vessel overpressure transient event is a main steam f
barrier in the prevention of uncontrolled release of fission products.
isolation valve closure with flux scram. This event was analyzed lt is essential that the integrity of this boundary be protected by within NEDC-32016P-1, " Power Uprate Safety Analysis For James cstablishing a pressure limit to be observed for all operating A. FitzPatrick Nuclear Power Plant," including Errata and Addenda coriditions and whenever there is irradiated fuel in the reactor Sheet No.1, dated January 1994, assuming 9 of the 11 SRVs were vessel.
operable with opening pressures less than or equal to 1179 psig.
l The resultant peak vessel pressure for the event was shown to be The pressure safety limit of 1,325 psig as measured by the vessel less than the ASME Code limit of 1375 psig (see current reload steam space pressure indicator is equivalent to 1,375 psig at the analysis for the reactor response to the main steam isolation valve lowest elevation of the Reactor Coolant System. The 1,375 psig closure with flux scram event).
value is derived from the design pressures of the reactor pressure vessel and reactor coolant system piping. The respective design A safety limit is applied to the Residual Heat Removal System pressures are 1250 psig at 575 *F for the reactor vessel,1148 psig (RHRS) when it is operating in the shutdown cooling mode. When et 568 *F for the recirculation suction piping and 1274 psig at 575 operating in the shutdown cooling mode, the RHRS is included in
- F for the discharge piping. The pressure safety limit was chosen the reactor coolant system.
l es the lower of the pressure transients permitted by the applicable
[
design codes: 1965 ASME Boiler and Pressure Vessel Code, Section ill for pressure vessel and 1969 ANSI B31.1 Code for the l
reactor coolant system piping. The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10 percent over dtsign pressure (110% x 1,250 = 1,375 psig) and the ANSI Code permits pressure transients up to 20 percent over the design pressure (120% x 1,150 = 1,380 psig). The safety limit pressure of 1,375 psig is referenced to the lowest elevation of the Reactor Coolant System.
i I
i Amendment No. SS, Si,134, la^, 217, 239 29 l
I
JAFNPP 3.1 BASES (cont'd)
The Control Rod Drive Scram System is designed so that The IRM high flux and APRM.115% power scrams all of the water which is discharged from the reactor by a provide adequate coverage in the startup and intermediate scram can be accommodated in the discharge piping.
ra.1ge. Thus, the IRM and APRM systems are required to Each scram discharge instrument volume accommodates be operable in the refuel and startup/ hot standby modes.
in excess of 34 gallons of water and is the low point in The APRM s120% power and flow referenced scrams l
the piping. No credit was taken for this volume in the provide required protection in the power range (reference design of the discharge piping as concems the amount of FSAR Section 7.5.7). The power range is covered only by water which must be accommodated during a scram.
the APRMs. Thus, the IRM system is not required in the rurt mode.
During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to The high reactor pressure, high drywell pressure, reactor the piping from the reactor could not be accommodated, low water level and scram discharge volume high level which would result in slow scram times or partial control scrams are required for startup and run modes of plant rod insertion. To preclude this occurrence, level detection operation. They are, therefore, required to be operational instruments have been provided in each instrument volume for these modes of reactor operation.
which alarm and scram the reactor when the volume of water reaches 34.5 gallons. As indicated above, there is The requirement to have the scram functions indicated in sufficient volume in the piping to accommodate the scram Table 3.1-1 operable in the refuel mode assures that without impairment of the scram times or amount of shifting to the refuel mode during reactor power operation insertion of the control rods. This function shuts the does not diminish the protection provided by the Reactor reactor down while sufficient volume remains to Protection System.
accommodate the discharged water and precludes the situation in which a scram would be required but not be Turbine stop valve closure occurs at 10 percent of valve able to perform its function adequately.
closure. Below 29% of rated reactor power, the scram l
signal due to turbine stop valve closure is bypassed A Source Range Monitor (SRM) System is also provided to I:ecause the flux and pressure scrams are adequate to supply additional neutron level information during startup protect the reactor.
but has no scram functions (reference paragraph 7.5.4 FSAR).
i l
Amendment No. 75,?3',207, 239 34 I
i JAFNPP 3.1 BASES (cont'd) f i
Turbine control valves fast closures initiates a scram C.
References based on pressure switches sensing electro-hydraulic i
control (EHC) system oil pressure. The switches are 1.
" James A. FitzPatrick Nuclear Power Plant located between fast closure solenoids and the disc dump SAFER /GESTR-LOCA Loss-of-Coolant Accident valves, and are set relative (500<P<850 psig) to the Analysis," NEDC-31317P, Revision 2, April 1993.
normal (EHC) oil pressure of 1,600 psig so that based on the small system volume, they can rapidly detect valve closure or loss of hydraulic pressure.
The requirement that the IRM's be inserted in the core j
when the APRM's read 2.5 indicated on the scale in the start-up and refuel modes assures that there is proper overlap in the neutron monitoring system functions and thus, that adequate coverage is provided for all ranges of reactor operation.
B.
The limiting transient which determines the required l
steady state MCPR limit depends on cycle exposure. The
[
l operating limit MCPR values as determined from the transient analysis in the current reload submittal for
[
various core exposures are specified in the Core Operating Limits Report (COLR).
l The ECCS performance analyses assumed reactor
?
operation will be limited to the MCPR value for each fuel I
type as described in Reference 1. The Technical l
Specifications limit operation of the reactor to the more conservative MCPR based on consideration of the limiting transient as specified in the COLR.
I t
l Amendment No. Ma, Si, ? ^^,1S2, ":dxf 5; "C '... ; f ^-f 3/1 S/^2, 239 i
35 t
l JAFNPP l
TABLE 3.1-1 (cont'd)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS l
Minimum No. of Mode in Which Function Operable Instrument Must Be Operable Total Number of Channels Per Instrument Channels i
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) 2 Reactor High Pressure s 1080 psig X
X
'X 4
A l
(Note 9) 2 Drywell High Pressure s2.7 psig X
X X
4 A
i (Note 16)
(Note 8) (Note 8) 2 Reactor Low Water 2:177 in. above TAF X
X X
4 A
i Level (Note 16) l f
3 High Water Level in s34.5 gallons per X
X X
8 A
Scram Discharge Volume Instrument Volume (Note 4)
L 4
Main Steam Line s10% valve closure X
8 A
i isolation Valve Closure (Note 6)
I 2
Turbine Control 500 < P < 850 psig X
4 A or C l
Valve Fast Closure Control oil pressure (Note 5) i between fast closure
[
solenoid and disc dump valve i
4 Turbine Stop s10% valve closure X
8 A or C Valve Closure (Notes 5 & 6) i Amendment No. 19,30,13,72,07,90,131,102,227, 239 i
41
[
I
JAFNPP TABLE 3.1-1 (cont'd)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS NOTES OF TABLE 3.1-1 (cont'd)
- 3. Action Statements:
A.
Insert all operable control rods within four hours.
B.
Reduce power level to IRM range and place Mode Switch in the Startup position within eight hours.
C.
Reduce power level to less than 29 percent of rated within four hours.
l 4.
Permissible to bypass, if the Reactor Mode Switch is in the Refuel or Shutdown position.
- 5. Bypassed when reactor power is less than 29 percent of rated power.
l
- 6. The design permits closure of any two lines without a scram being initiated.
f
- 7. When the reactor is subcritical and the reactor water temperature is less than 212*F, only the following trip functions need to be operable:
A.
Mode Switch in Shutdown.
i B.
Manual Scram.
C.
High Flux IRM D.
Scram Discharge Volume High Level when any control rod in a control cell containing fuel is not fully inserted.
E.
APRM 15% Power Trip.
8.
Not required to be operable when primary coatainment integrity is not required.
9.
Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
I
- 10. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 11 LPRM inputs of the normal
~
complement.
i
- 11. (Deleted)
Amendment No. 19, 62, Si, S7, S0, 72, 71,109, * ' 7,159,162, 207, 227, 239 43
i JAFNPP i
3.3 and 4.3 BASES (cont'd) l
" full out" position during the performance of SR 3.
The Rod Worth Min 3mizer (RWM) restricts the order of i
4.3.A.2.a. This Frequency is acceptable, considering the control rod withdrawal and insertion to be equivalent to low probability thet a control rod will become uncoupled the Banked Position Withdrawal Sequence (BPWS). These when it is not being moved, and operating experience sequences are established such that the drop of any related to uncoupling events.
in-sequence control rod from the fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in a peak 2.
The control rod housing support restricts the outward fuel enthalpy in excess of 280 cal /gm.. An enthalpy of movement of a control rod to less than 3 in. in the 280 cal /gm is well below the level at which rapid fuel l
l extremely remote event of a housing failure. The amount dispersal could occur (i.e. 425 cal /gm.). Primary system
(
of reactivity which could be added t y this small amount damage in this accident is not possible unless a i
r of rod withdrawal, which is less than a normal single significant amount of fuelis rapidly dispersed. Ref.
withdrawal increment, will not contribute to any damage Subsections 3.6.6, 7.7.4.3 and 14.6.1.2 of the FSAR, to the Primary Coolant System. The design basis is given NEDE-24011-P-A-13, August 1996 and NEDO-10527 l
in subsection 3.8.2 of the FSAR, and the safety including supplements 1 and 2 to NEDO-10527.
evaluation is given in subsection 3.8.4. This support is not required if the Reactor Coolant System is at in performing the function described above, the RWM is i
atmospheric pressure since there would then be no not required to impose any restrictions at core power driving force to rapidly eject a drive housing.
levels in excess of 10% of rated. Material in the cited Additionally, the support is not required if all control rods references shows that it is impossible to reach 280 l
are fully inserted and if an adequate shutdown margin calories per gram in the event of a control rod drop l
with one control rod withdrawn has been demonstrated, occurring at power greater than 10%, regardless of the since the reactor would remain subcritical even in the rod pattern. This is true for all normal and abnormal event of complete ejection of the strongest control rod.
patterns including those which maximize the individual control rod worth.
3 i
i i
I I
Amendment No. 30,155,193, 239 f
100
i JAFNPP L
3.3 and 4.3 BASES (cont'd) j 5.
The Rod Block Monitor (RBM) is designed to automatically C.
Scram insertion Times prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level The Control Rod System is designated to bring the reactor-operation. Two channels are provided, and one of these may subcritical at a rate fast enough to prevent fuel damage; be bypassed from the console for maintenance and/or testing.
i.e., to prevent the MCPR from becoming less than the t
Tripping of one of the channels will block erroneous rod Safety Limit. Scram insertion time test criteria of Section
[
withdrawal soon enough to prevent fuel damage.
3.3.C.1 were used to generate the generic scram reactivity curve shown in NEDE-24011-P-A-13, August This system backs up the operator who withdraws control 1996. This generic curve was used in analysis of 7
rods according to written sequences. The specified non-pressurization transients to determine MCPR limits.
t restrictions with one channel out of service conservatively Therefore, the required protection is provided, assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.
The numerical values assigned to the specified scram A limiting control rod pattern is a pattern which results in the performance are based on the analysis of data from other core being on a thermal hydraulic limit (e.g., MCPR limit).
BWR's with control rod drives the same as those on j
During use of such pattems, it is judged that testing of the JAFNPP.
j RBM System prior to withdrawal of such rods to assure its operability will assure that improper withdraw does not occur.
The occurrence of scram times within the limits, but it is the responsibility of the Reactor Engineer to identify these significantly longer than the average, should be viewed as I
limiting pattems and the designated rods either when the an indication of a systematic problem with control rod patterns are initially established or as they develop due to the drives, especially if the number of drives exhibiting such occurrence of inoperable control rods in other than limiting scram times exceeds eight, the allowable number of j
patterns.
inoperable rods.
o i
i 239 l
Amendment No. '" '^
'n eo e,
.re
.e, 102 l
D
JAFNPP 3.5 (cont'd) 4.5 (cont'd)
DELETED C.
HIGH PRESSURE COOLANT INJECTION (HPCI SYSTEM)
C.
HIGH PRESSURE COOLANT INJECTION (HPCI SYSTEM)
Surveillance of HPCI System shall be performed as follows provided a reactor steam supply is available. If steam is not available at the time the surveillance test is scheduled to be performed, the test shall be performed within 10 days of continuous operation from the time steam becomes 1.
The HPCI System shall be operable whenever the available.
reactor pressure is greater than 150 psig and reactor coolant temperature is greater than 212*F and 1.
HPCI System testing shall be as specified in irradiated fuel is in the reactor vessel, except es 4.5.A.I.a, b, c, d, f, and g except that the HPCI specified below:
pump shall deliver at least 4,250 gpm against a system head corresponding to a reactor vessel pressure of 1,195 psig to 150 psig.
I i
t Amendment No. 40,95,107, 239 117
?
JAFNPP 3.5 (cont'd) 4.5 (cont'd)
The RCIC pump shall deliver at least 400 gpm for a system head corresponding to a reactor pressure of 1.195 psig to I
150 psig.
i 2.
When it is determined that the RCIC System is inoperable at a time when it is required to be operable, the HPCI System shall be verified to be operable immediately and daily thereafter.
1 i
I l
i I
l Amendment No. 10, ' 18, 239 i
121a 6
P
-x
JAFNPP 3.5 BASES A.
Core Sorav System and Low Pressure Coolant iniection (LPCI)
Core spray distribution has been shown, in full scale tests of I
Mode of the RHR System systems similar in design to that of the FitzPatrick Plant, to exceed the minimum requirements by at least 25 percent. In addition, This specification assures that adequate emergency cooling cooling effectiveness has been demonstrated at less than half the capability is available whenever irradiated fuel is in the reactor rated flow in simulated fuel assemblies with heater rods to vessel.
duplicate the decay heat characteristics of irradiated fuel. The accident analysis is additionally conservative in that no credit is The loss-of-coolant analysis is referenced and described in taken for spray coolant entering the reactor before the internal General Electric Topical Report NEDE-24011-P-A-13, August pressure has fallen to 113 psi above primary containment
- 1996, pressure.
The limiting conditions of operation in Specifications 3.5.A.1 The LPCI mode of the RHR System is designed to provide through 3.5.A.6 specify the combinations of operable emergency cooling to the core by flooding in the event of a subsystems to assure the availability of the minimum cooling loss-of-coolant accident.
These subsystems are completely i
systems. No single failure of ECCS equipment occurring during independent of the Core Spray System; however, they function in
[
a loss-of-coolant accident under these limiting conditions of combination with the Core Spray System to prevent excessive fuel operation will result in inadequate cooling of the reactor core.
clad temperature. The LPCI mode of i
L i
Amendment No. 48, 'ie,
- 10, 239 125
[
1
JAFNPP 3.6 (cont'd) 4.6 (cont'd)
B.
Deleted B.
Deleted l
C.
Coolant Chemistry C.
Coolant Chemistry 1.
The reactor coolant system radioactivity concentration in 1.
a.
A sample of reactor coolant shall be taken at least l
water shall not exceed the equilibrium value of 0.2 every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and analyzed for gross gamma pCi/gm of dose equivalent 1-131. This limit may be activity.
exceeded, following a power transient, for a maximum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. During this iodine activity transient the iodine b.
Isotopic analysis of a sample of reactor coolant shall concentrations shall not exceed the equilibrium limits by be made at least once/ month.
more than a factor of 10 whenever the main steamline isolation valves are open. The reactor shall not be c.
A sample of reactor coolant shall be taken prior to j
operated more than 5 percent of its annual power startup and at 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> intervals during startup and
~
operation under this exception to the equilibrium limits. If analyzed for gross gamma activity.
the iodine concentration exceeds the equilibrium limit by more than a factor of 10, the reactor shall be placed in a d.
During plant steady state operation and following an cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
offgas activity increase (at the Steam Jet Air Ejectors) of 10,000 pCi/sec within a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period or a power level change of 2:20 percent of full rated i
power /hr reactor coolant samples shall be taken and analyzed for gross gamma activity. At least three samples will be taken at 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> intervals. These i
sampling requirements may be omitted whenever the equilibrium l-131 concentration in the reactor coolant is less than 0.007 pCi/ml.
l i
i Amendment No.
- 79,1"^,199, 239 139
JAFNPP 3.6 (cont'd) 4.6 (cont'd)
E.
Safetv/ Relief Valves E.
Safetv/ Relief Valves
- 1. During reactor power operating conditions and prior to
- 1. At least 5 of the 11 safety / relief valves shall be bench l
startup from a cold condition, or whenever reactor coolant checked or replaced with bench checked valves every 24 pressure is greater than atmosphere and temperature months. All valves shall be tested every 48 months. The greater than 212 F, the safety mode of at least 9 of 11 testing shall demonstrate that each valve tested actuates safety / relief valves shall be operable. The Automatic at 1145 psig 13%. Following testing, lift settings shall be Depressurization System valves shall be operable as 1145 psig 11%.
required by specification 3.5.D.
i r
i i
Amendment No. 13, 28, SS, 70,130, 131,195, 204, 217, 219, 229, 239 i
142a
{
3.6 and 4.6 BASES (cont'd) i The expected neutron fluence at the reactor vessel wall can be The RTer for the remainder of the vessel is 40 F.
determined at any point during plant life based on the linear relationship between the reactor thermal power output and the The actual shift in the RT., of the vessel material will be corresponding number of neutrons produced. Accordingly, established periodically by removing and evaluating flux neutron flux wires were removed from the reactor vessel with monitoring surveillance capsules in accordance with ASTM E t
the surveillance specimens to establish the correlation at the 185-82 and 10 CFR 50, Appendix H. The evaluation findings capsule location by experimental methods. The flux and recommendations of Regulatory Guide 1.99 Revision 2 will distribution at the vessel wall and 1/4 thickness (1/4T) depth provide the basis for revising Figure 3.6-1 curves A, B and C was analytically determined as a function of core height and for operation of the plant. The first surveillance capsule azimuth to establish the peak flux location in the vessel and containing test specimens was withdrawn in April,1985 after the lead factor of the surveillance specimens.
6 EFPY. The test specimens removed were tested according to ASTM E 185-82 and the results are in GE report MDE Regulatory Guide 1.99, Revision 2 is used to predict the shift 0386. The NRC approved schedule for subsequent specimen t
in RT y as a function of fluence in the reactor vessel beltline withdrawal is located in the updated FSAR (Section 4.2.7).
region. An evaluation of the irradiated surveillance specimens, which were withdrawn from the reactor in April,1985 (6 Figure 3.6-1 is comprised of three parts: Part 1, Part 2, and EFPY), shows a shift in RT.y less than that predicted by Part 3. Parts 1, 2,' and 3 establish the pressure-temperature i
Regulatory Guide 1.99, Revision 2.
limits for plant operations through 12,14, and 16 Effective Full Power Years (EFPY) respectively. The appropriate figure Operating limits for the reactor vessel pressure and and the pressure-temperature curves are dependent on the temperature during normal heatup and cooldown, and during number of accumulated EFPY. Figure 3.6-1, Part 1 is for in-service hydrostatic and leak testing were established using operation through 12 EFPY, Figure 3.6-1, Part 2 is for 10 CFR 50 Appendix G, May,1983 and Appendix G of the operation at greater than 12 EFPY through 14 EFPY, and Summer 1984 Addenda to Section 111 of the ASME Boiler and Figure 3.6-1, Part 3 is for operation at greater than 14 EFPY l
Pressure Vessel Code. These operating limits assure that the through 16 EFPY. The curves contained in Figure 3.6-1 are vessel could safely accommodate a postulated surface flaw developed from the General Electric Report DRF 137-0010, having a depth of 0.24 inch at the flange-to-vessel junction,
" Implementation of Regulatory Guide 1.99, Revision 2 for the i
and one-quarter of the material thickness at all other reactor James A. FitzPatrick Nuclear Power Plant," dated June,1989.
vessel locations and discontinuity regions. For the purpose of I
setting these operating limits, the reference temperature, Figure 3.6-1 curve A establishes the minimum temperature for l
RT.1, of the vessel material was estimated from impact test hydrostatic and leak testing required by the ASME Boiler and j
data taken in accordance with the requirements of the Code to Pressure Vessel Code,Section XI. Test pressures for which the vessel was designed and manufactured (1965 in-service hydrostatic and leak testing are a function of the
{
Edition including Winter 1966 addenda). The RTer values for testing temperature and the component material. Accordingly, the reactor vessel flange region and for the reactor vessel shell the maximum hydrostatic test pressure will be 1.1 times the beltline region are 30*F, based on fabrication test reports.
operating pressure or about 1,144 psig.
l l
l i
t l
Amendment No.
- 13,158,199, 239 f
l 147
i JAFNPP 3.6 and 4.6 BASES (cont'd) l l
B.
Deleted annunciating at appropriate concentration levels such that i
sampling for isotopic analysis can be initiated. The design i
C.
Coolant Chemistry details of such a system must be submitted for evaluation and accepted by the Commission prior to its implementation and I
A radioactivity concentration limit of 20 pCi/ml total iodine incorporation in these Technical Specifications.
can be reached if the gaseous effluents are near the limit as set forth in Radiological Effluent Technical Specification Since the concentration of radioactivity in the reactor coolant is Section 3.2.a if there is a failure or a prolonged shutdown of not continuously measured, coolant sampling would be l
the cleanup demineralizer.
ineffective as a means to rapidly detect gross fuel element failures. However, some capability to detect gross fuel element in the event of a steam line rupture outside the drywell, a failures is inherent in the radiation monitors in the offgas system more restrictive coolant activity level of 0.2 pCi/gm of dose and on the main steam knes.
i equivalent 1-131 was assumed. With this coolant activity i
level and adverse metecrological conditions, the calculated Materials in the Reactor Coolant System are primarily 304 i
radiological dose at the site boundary would be less than 30 stainless steel ar:d Zircaloy fuel cladding. The reactor water i
rem to the thyroid. The reactor water sample will be used to chemistry limits are established to prevent damage to these assure that the limit of Srecification 3.6.C is not exceeded.
materials. Limits are placed on chloride concentration and The total radioactive iodiae activity would not be expected to conductivity. The most important limit is that placed on change rapidly over a pariod of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. In addition, the chloride concentration to prevent stress corrosion cracking of trend of the stack ofigas release rate, which is continuously the stainless steel. The attached graph, Fig. 4.6-1, illustrates monitored, is a good indicator of the trend of the iodine the results of tests on stressed 304 stainless steel specimens.
activity in the reactor coolant. Also during reactor startups Failures occurred at concentrations above the curve; no failures and large pwer changes which could affect iodine levels, occurred at concentrations below the curve. According to the
[
samples of reactor coolant shall be analyzed to insure iodine data, allowable chloride concentrations could be set several concentrations are below allowable levels. Analysis is orders of magnitude above the established limit, at the oxygen required whenever the 1-131 concentration is within a factor concentration (0.2-0.3 ppm) experienced during power j
of 100 of its allowable equilibrium value. The necessity for operation. Zircaloy does not exhibit similar stress corrosion continued sampling following power and offgas transients will failures.
i be reviewed within 2 years of initial plant startup.
i However, there are various conditions under which the
[
The surveillance requirements 4.6.C.1 may be satisfied by a dissolved oxygen content of the reactor coolant water could be continuous monitoring system capable of determining the total higher than 0.2-0.3 ppm, such as refueling, reactor startup, and iodine concentration in the coolant on a real time basis, and hot standby. During these periods with steaming rates less i
l Amendment No.
- 79, t a^,239 4
149 l
m i
JAFNPP 3.6 and 4.6 BASES (cont'd)
E.
Safetv/ Relief Valves The safety / relief valves (SRVs) have two modes of operation; with the HPCI and RCIC turbine overspeed systems and the f
the safety made or the relief mode. In the safety mode (or Mark I torus loading analyses. Based on safety / relief valve l
spring mode of operation) the spring loaded pilot valve opens testing experience and the analysis referenced above, the I
when the steam pressure at the valve inlet overcomes the spring safety / relief valves are bench tested to demonstrate that
[
force holding the pilot valve closed. The safety mode of in-service opening pressures are within the nominal pressure l
operation is required during pressurization transients to ensure setpoints 13% and then the valves are returned to servce i
vessel pressures do not exceed the reactor coolant pressure with opening pressures at the nominal setpoints 11 %. In this safety limit of 1,375 psig, manner, valve integrity is maintained from cycle to cycle.
In the relief mode the spring loaded pilot valve opens when the The analyses with NEDC-32016P-1, including Errata and t
spring force is overcome by nitrogen pressure which is provided Addenda Sheet No.1, dated January 1994, also provide the i
to the valve through a solenoid operated valve. The solenoid safety basis for which 2 SRVs are permitted inoperable during operated valve is actuated by the ADS logic system (for those continuous power operation. With more than 2 SRVs SRVs which are included in the ADS) or manually by the inoperable, the margin to the reactor vessel pressure safety -
operator from a control switch in the main control room or at limit is significantly reduced, therefore, the plant must enter a i
the remote ADS panel. Operation of the SRVs in the relief mode cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> once more than 2 SRVs are
[
for the ADS is discussed in the Bases for Specification 3.5.D.
determined to be inoperable. (See reload evaluation for the current cycle).
Experiences in safety / relief valve testing have shown that failure or deterioration of safety / relief valves can be adequately A manuoi actuation of each SRV is performed to demonstrate detected if at least 5 of the 11 valves are bench tested once that the valves are mechanically functional and that no i
every 24 months so that all valves are tested every 48 months.
blockage exists in the valve discharge line. Valve opening is Furthermore, safety / relief valve testing experience has confirmed by monitoring the response of the turbine bypass L
demonstrated that safety / relief valves which actuate within valves and the SRV acoustic monitors. Adequate reactor 13% of the design pressure setpoint are considered operable steam dome pressure must be available to avoid damaging the (see ANSI /ASME OM-1-1981). The safety bases for a single valve. Adequate steam flow is required to ensure that reactor nominal valve opening pressure of 1145 psig are described in pressure can be maintained dunng the test. Testing is NEDC-32016P-1, " Power Uprate Safety Analysis for James A.
performed in the RUN mode to reduce the risk of a reactor FitzPatrick Nuclear Power Plant," Including Errata and Addenda scram in response to small pressure fluctuations which may I
Sheet No.1, dated January 1994. The single nominal setpoint occur while opening and reclosing the valves.
is set below the reactor vessel design pressure (1250 psig) per j
the requirements of Article 9 of the ASME Code - Section Ill, Low power physics testing and reactor operator training with I
Nuclear Vessels. The setting of 1145 psig preserves the safety inoperable components will be conducted only when the j
margins associated safety / relief valves are I
r Amendment No. 13, 131, 217, 219, 229, 239 152 l
JAFNPP 3.7 (cont'd) 4.7 (cont'd)
(2)
During testing which adds heat to the suppression pool, the water temperature shall not exceed 10*F above the normal power operation limit specified in l
(1) above. In connection with such testing, the pool temperature must be reduced to below the normal power operation limit specified in (1) above within 24 l
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i (3)
The reactor shall be scrammed from any operating condition if the pool temperature reaches 110*F.
Power operation shall not be resumed until the pool temperature is reduced below the normal power operation limit specified in (1) above.
i
?
(4)
During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200 psig at normal cooldown rates if the pool t
temperature reaches 120*F.
2.
Primary containment integrity shall be maintained at all times 2.
a.
Perform required visual examination and leakage rate when the reactor is critical or when the reactor water testing of the Primary Containment in accordance temperature is above 212*F, and fuel is in the reactor with the Primary Containment Leakage Rate Testing vessel, except while performing low power physics tests at Program.
atmospheric pressure at power levels not to exceed 5 MWt.
b.
Demonstrate leakage rate through each MSIV is s 11.5 scfh when tested at 2 25 psig. The testing I
frequency is in accordance with the Primary Containment Leakage Rate Testing Program.
t c.
Once per 24 months, demonstrate the leakage rate of i
10AOV-68A,8 for the Low Pressure Coolant injection 7
system and 14AOV-13A,B for the Core Spray system i
to be less than 11 scfm per valve when pneumatically tested at a 45 psig at ambient temperature, or less than 10 gpm per valve if hydrostatically tested at 2 i
1,035 psig at ambient temperature.
l l
l 1
i Amendment No. 1 S, 231, 239 l
166
JAFNPP 3.7 BASES A.
Primary Containment The integrity of the primary containment and operation of the The pressure suppression pool water provides the heat sink for Emergency Core Cooling Systems in combination limit the the Reactor Coolant System energy release following a offsite doses to values less than those specified in 10 CFR postulated rupture of the system. The pressure suppression 100 in the event of a break in the Reactor Coolant System chamber water volume must absorb the associated decay and piping. Thus, containment integrity is required whenever the structural sensible heat released during reactor coolant system potential for violation of the Reactor Coolant System integrity blowdown from 1,040 psig.
I exists. Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An Since all of the gases in the drywell are purged into the exception to the requirement to maintain primary containment pressure suppression chamber air space during a loss of integrity is allowed during core loading and during low power coolant accident, the pressure resulting from isothermal physics testing when ready access to the reactor vessel is compression plus the vapor pressure of the liquid must not required. There will be no pressure on the system at this time, exceed 56 psig, the suppression chamber design pressure.
which will greatly reduce the chances of a pipe break. The The design volume of the suppression chamber (water and air) reactor may be taken critical during this period, however, was obtained by considering that the total volume of reactor restrictive operating procedures and operation of the RWM in coolant to be condensed is discharged to the suppression accordance with Specification 3.3.B.3 minimize the probability chamber and that the drywell volume is purged to the of an accident occurring. Procedures in conjunction with the suppression chamber (updated FSAR Section 5.2).
Rod Worth Minimizer Technical Specifications limit individual control worth such that the drop of any in-sequence control rod would not result in a peak fuel enthalpy greater than 280 calori':s/gm. In the unlikely event that an excursion did occur, the reactor building and Standby Gas Treatment System, which shall be operational during this time, offers a sufficient barrier to keep offsite doses well within 10 CFR 100.
f i
i Amendment No. 1 S,155,190, 239 187
. - - -. - -. ~ - - -
l i
JAFNPP 3.7 BASES (cont'd)
Using the minimum or maximum torus water level (which are temperature. Therefore, complete condensation is assured based on downcomer submergence levels where 13.88 feet during a LOCA because the maximum pool temperature
~
above the bottom of the torus is 0.005 feet higher than the (141 *F) is less than the 170*F temperature seen during the
[
minimum submergence of 51.5 inches and 14.00 feet above Bodega Bay tests.
the bottom of the torus is equivalent to the maximum submergence of 53 inches assumed in containment ans!yses)
For an initial maximum torus water temperature of 95aF, containment pressure during the design basis accident is assuming the worst case complement of containment cooling approximately 45 psig which is below the design of 56 psig.
pumps (one LPCI pump and two RHR service water pumps),
The minimum downcomer submergence of 51.5 inches results containment pressure is required to maintain adequate net l
in a minimum torus water volume of approximately 105,900 positive suction head (NPSH) for the core spray and LPCI' 8
feet. The majority of the Bodega tests (9) were run with a pumps.
submerged length of 4 feet and with complete condensation.
1 Thus, with respect to downcomer submergence, this Limiting suppression pool temperature to 105 F during RCIC, l
l specification is adequate. Additional JAFNPP specific analyses HPCI, or relief valve operation, when decay heat and stored done in connection with the Mark l Containment-Suppression energy are removed from the primary system by discharging l
[
l Chamber integrity Program indicate the adequacy of the reactor steam directly to the torus assures adequate margin for t
specified range of submergence to ensure that dynamic forces a potential blowdown any time during RCIC, HPCI, or relief l
associated with pool swell do not result in overstress of the valve operation.
j torus or associated structures. Level instrumentation is l
provided for operator use to maintain downcomer Experiments indicate that unacceptably high dynamic I
submergence within the specified range.
containment loads may result from unstable condensation when suppression pool water temperatures are high near SRV The maximum temperature at the end of blowdown tested discharges. Action statements limit the maximum pool during the Humboldt Bay (10) and Bodega Bay tests was temperature to assure stable condensation. These actions i
170 F, and this is conservatively taken to be the limit for include: limiting the maximum pool temperature of 95'F I
complete condensation of the reactor coolant, althe"gh during normal operation; initiating a reactor scram if during a j
condensation would occur for temperatures above 170*F.
transient (such as a stuck open SRV) pool temperature l
exceeds 110*F; and depressurizing the reactor if pool Containment analyses predict a 46*F increase in pool water temperature exceeds 120*F. T-quenchers diffuse steam temperature, after complete LOCA blowdown. These analyses discharged from SRVs and promote stable condensation. The assumed an initial suppression pool water temperature of 95 F presence of T-quenchers and compliance with these action i
and a rated reactor power of 2536 MWt. LOCA analyses in statements assure that stable condensation will occur and Section 14.6 of the FSAR also assume an initial 95 F pool containment loads will be acceptable.
j l
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l 239 i
Amendment No. 16, 3S,1SS, ? 01,197, 188 i
m _ _. _.
JAFNPP l
3.7 BASES (cont'd)
NEDC-24361P (August 1981) summarizes analyses performed significant heat addition, the temperature trends will be closely to predict pool temperatures and containment loads during followed so that appropriate action can be taken. There are plant transients using these temperature limits at a power level alarms at applicable limits to provide further assurance of l
of 2535 MWt (104% of rated). NEDC-24361P also appropriate action. The requirement for an external visual l
substantiates the acceptability of the plant design using the examination following any event where potentially high local pool limits of NUREG-0661. NEDO-30832 (December loadings could occur provides assurance that no significant l
1984) shows that SRV condensation loads are low compared damage was encountered. Particular attention should be l
to other design loads for plants with T-quenchers. NEDO-focused on structural discontinuities in the vicinity of the relief 30832 describes why local pool temperatures need not be valve discharge since these are expected to be the points of l
analyzed at a rated power level of 2536 MWt.
highest stress.
In addition to the limits on temperature of the suppression pool if a loss-of-coolant accident were to occur when the reactor water, operating procedures define the action to be taken in water temperature is below 330'F, the containment pressure the event a relief valve inadvertently opens or sticks open.
will not exceed the 56 psig design pressure, even if no These procedures include: (1) use of all available means to condensation were to occur. The maximum allowable pool close the valve, (2) initiate suppression pool cooling, (3) temperature, whenever the reactor is above 212*F, shall be i
initiate reactor shutdown, and (4) if other relief valves are used governed by this i
to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and
[
uniformity of energy insertion to the pool.
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally 1
changes very slowly and monitoring these parameters daily is i
sufficient to establish any temperature trends. By requiring r
the suppression pool temperature to be verified as within applicable limits every 5 minutes as well as continuously
[
recorded (the operator can log temperature during verification i
if continuous recording is not available) during periods of i
I Amendment No.1S,197, 239 i
188a
~-. - - -
i JAFNPP i
4.7 BASES i
A.
Primary Containment Design basis accidents were evaluated as discussed in Section 14.6 of the FSAR and the power uprate safety The water in the suppression chamber is used only for evaluation, Reference 18. The whole body and thyroid
[
cooling in the event of an accident; i.e., it is not used for doses in the control room, low population zone (LPZ) and normal operation; therefore, a daily check of the site boundary meet the requirements of 10 CFR Parts 50 i
temperature and volume is adequate to assure that and 100. The technical support center (TSC), not i
adequate heat removal capability is present.
designed to these licensing bases, was also analyzed. The I
whole body and thyroid dose acceptance criteria used for The primary containment preoperational test pressures are the main control room are met for the TSC when initial i
based upon the calculated primary containment pressure access to the TSC and occupancy of certain areas in the response corresponding to the design basis loss-of-coolant TSC is restricted by administrative control. The LOCA accident. The peak drywell pressure would be about 45 dose evaluations, References 19, 20, and 21, assumed:
l psig which would rapidly reduce to 27 psig within 30 sec.
the primary containment leak rate was 1.5 volume percent i
following the pipe break. Following the pipe break, the per day; source term releases were in accordance with suppression chamber pressure rises to 26 psig within 30 TID-14844 and Regulatory Guide 1.3, and were consistent i
sec, equalizes with drywell pressure and thereafter rapidly with the Standard Review Plan; and the standby gas j
decays with the drywell pressure decay (14).
treatment system filter efficiency was 99% for halogens.
+
These doses are also based on the l
The design pressure of the drywell and suppression chamber is 56 psig(15). The design basis accident leakage rate is 0.5 percent / day at a pressure of 45 psig.
[
As pointed cut above, the drywell and suppression chamber pressure following an accident would equalize fairly rapidly. Based on the primary containment pressure
}
response and the fact that the drywell and suppression r
chamber function as a unit, the primary containment will be tested as a unit rather than the individual components separately.
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239 Amendment No.
j 193 l
JAFNPP l
(A)
ROUTINE REPORTS (Continued) l 4.
CORE OPERATING LIMITS REPORT 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:
The Average Planar Linear Heat Generation Rates (APLHGR) of Specification 3.5.H; I
The Minimum Critical Power Ratio (MCPR) and MCPR low flow adjustment factor, K, of Specifications 3.1.B and 4.1.E:
i The Linear Heat Generation Rate (LHGR) of Specification 3.5.l; 1
The Reactor Protection System (RPS) APRM flow biased trip settings of i
Table 3.1-1; The flow biased APRM and Rod Block Monitor (RBM) rod block settings of Table 3.2-3; and The Power / Flow Exclusion Region of Specification 3.5.J.
and shall be documented in the Core Operating Limits 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-A-13, August 1996.
I
- 2. " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR - LOCA Loss-of Coolant Accident Analysis," NEDC-31317P, Revision 2, April 1993.
i
- 3. "BWR Owners' Group Long-term Stability Solutions Licensing l
Methodology," NEDO-31960-A, June 1991.
- 4. "BWR Owners' Group Long-term Stability Solutions Licensing 1
Methodology," NEDO-31960-A, Supplement 1, March 1992.
l l
i Amendment No.162, 23S, 239 254c
i i
i JAFNPP f
7.0 REFERENCES
t I
(1)
E. Janssen, " Multi-Rod Burnout at Low Pressure," ASME Paper (11) Section 5.2 of the FSAR.
62-HT-26, August 1962.
(12) TID 20583, " Leakage Characteristics of Steel Containment (2)
K.M. Backer, " Burnout Conditions for Flow of Boiling Water in.
Vessel and the Analysis of Leakage Rate Determinations."
i Vertical Rod Clusters," AE-74 (Stockholm, Sweden), May 1962.
(13) Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program", dated September 1995.
(3) FSAR Section 11.2.2.
(14) Section 14.6 of the FSAR.
(4) FSAR Section 4.4.3.
i (15) ASME Boiler and Pressure Vessel Code, Nuclear Vessels, I
(5) 1.M. Jacobs, "ReliabDity of Engineered Safety Features as a Section Ill. Maximum allowable internal pressure is 62 pseg.
t Function of Testing Frequency," Nuclear Safety, Vol. 9, No. 4, j
July-August 1968, pp 310-312.
(1ft) 10 CFR Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B -
(6) Deleted Performance Based Requirements", Effective Date October 26, 1995 t
(7) 1.M. Jacobs and P.W. Mariott, APED Guidelines for Determining Safe Test Intervals and Repair Times for (17) Deleted
{
Engineered Safeguards - April 1969.
l (18) General Electric Report NEDC-32016P-1, " Power Uprate I
(8) Bodega Bay Preliminary Hazards Report, Appendix 1, Docket Safety Analysis for James A. FitzPatrick Nuclear Power Plant,"
50-205, December 28,1962.
April 1993 (proprietary), including Errata and Addenda Sheet No.1, dated January 1994.
(9)
C.H. Robbins, " Tests of a Full Scale 1/48 Segment of the Humbolt Bay Pressure Suppression Containment,"
(19) James A. FitzPatrick Calculation JAF-CALC-RAD-OOO23, Rev.
GEAP-3596, November 17,1960.
O, " Power Uprate Program - Technical Support Center Post-
[
Accident Radiological Habitability Study," August 1996.
l (10) " Nuclear Safety Program Annual Progress Report for Period
[
l Ending December 31,1966, ORNL-4071."
(20) James A. FitzPatrick Calculation JAF-CALC-RAD-OOO42, Rev.
O, " Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration," September 1995.
Amendment No. "^ '" ' 239 285 I
JAFNPP l
7.0 REFERENCES
(continued)
(21) James A. FitzPatrick Calculation JAF-CALC-RAD-OOO48, Rev. O, " Power Uprate Project - Radiological impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents," May 1996 (22) General Electric Report GE-NE-187-45-1191,
" Containment Systems Evaluation for the James A.
FitzPatrick Nuclear Power Plant," November 1991 (proprietary).
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Amendment No. 239 285a 1
-.