ML19321B159
| ML19321B159 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/11/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19321B158 | List: |
| References | |
| NUDOCS 8007280249 | |
| Download: ML19321B159 (29) | |
Text
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j$j UNITED STATES 8
NUCLEAR PEGULATORY COMMISSION c(
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^I.I WASHINGTON, D. C. 20555 POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING i.ICENSE Amendment No. 49 License No. DPR-59' l.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Power Authority of the State of New York (the licensee) dated March 4, 1980, 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 ap;.lication, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such ictivities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment wil1 not be inimical to the common 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.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-59 is hereby amended to resd as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 49, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
g* %Q
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Thomas-A. Ippolito, Chief Operatir.3 Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: July 11,1980 n
I
ATTACHMENT TO LICENSE AMENDMENT N0. 49 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
Remove Insert 6
. 6 15 15 18 18 20 20 29 29 30 30 31 31 35 35 43 43 58 58 72 72 73 73 95 95 96 96 102 102 103 103 108 108 123 123 124 124 125 125 130 130 135e 135f 2:5 245 247 247 and 247a
JAFNPP surveillance tests, checks, calibrations, and V.
Electrically Disarmed Control Rod q) examinations shall be performed within the specified surveillance intervals. These intervals To disarm a rod drive electrically, the four may be adjusted + 25 percent. The interval as amphenol type plug connectors are removed pertaining to instrument and electric surveillance from the drive insert and withdrawal shall never exceed one operating cycle. Ir cases solenoids rendering the rod incapable of where the elapsed interval has exceeded 100 withdrawal. This procedure is equivalent percent of the specified interval, the next to valving out the drive and is preferred.
surveillance interval shall cammence at the end Electrical disarming does not eliminate of the original specified interval.
position indication.
?!
?
I U.
Thermal Parameters W.
High Pressure Water Fire Protection System 1.
Minimum critical power ratio (MCPR)-Ratio The High Pressure Water Fire Protection of that power in a fuel assmelby which is System consists of: a water source and calculated to cause some point in that fuel pumps: and distribution system piping with assembly to experience boiling transition associated post indicator valves (isolation l
to the actual assembly operating power as valves). Such valves include the yard calculated by application of the GEXL hydrant curb valves and the first valve
[
correlation (Reference NEDE-10958).
ahead of the water flow alarm device on each sprinkler or water spray subsystem.
2.
Fraction of Limiting Power Density - The ratio of the linear heat generation rate X.
Staggered Test Basis (LHGR) existing at a given location to the j
design LHGR for that bundle type. Design A Staggered Test Basis shall consist of:
LHGR's are 18.5 KW/ft for 7x7 bundles and l
13.4 KW/ft for 8x8, 8x8R and P8x8R bundles.
a.
A test schedule for a systems, sub-systems, trains or other designated 3.
Maximum Fraction of Limiting Power Density -
components obtained by dividing the The Maximum Fraction of Limiting Power specified test interval into n equal Density (MFLPD) is the highest value exist-subintervals.
ing in the core of the Fraction of Limiting Power Density (FLPD).
b.
The tes' ting of one system, subsystem, train or other designated component I
4.
Transition Boiling - Transition boiling means at the beginning of each subinterval, the boiling region between nucleate and film boiling. Transition boiling is the region in which both nucleate and film boiling occur intermittently with neither type being com-plately stable.
Amendment No. 49 6
JAFNPP I.
2.1 BASES 2.1 EVEL CIADDING IWFEGRITY
,e The abnormal operational'transiente appli-tool for evaluating reactor dynamic performance.
cable to operation of the FitzPatrick Unit Results obtained from a General Electric boiling have been analyzed throughout the spectrum water reactor have been compared with predictions i
of planned operating conditions up to the made by the model. ne comparisons and results f
thermal power condition of 2535 MWt.
The are sumunarized in Reference 1.
analyses were based upon plant operation in accordance with the operating map given in The absolute value of the void reactivity coefficient
~
Figure 3.7-1 of the FSAR.
In addition, 2436 used in the analysis is conservatively estimated to is the licensed maxistan power level of Fitz-be about 254 greater than the nominal maximum value Patrick, and this represents the maximum expected to occur during the core lifetime. The steady-state power which shall not knowingly scram worth used has been derated to be equivalent be exceeded.
to approximately 80% of the total scram worth of the control rods. he scram delay time and rate of rod Conservatism is incorporated in the transient insertion allowed by the analyses are conservatively analyses in estimating the controlling factors, set equal to the longest delay and slowest insertion such as void reactivity, coefficient, control rate acceptable by Technical Specifications. Active rod scram worth, scram delay time, peaking coolant flow is equal to 88% of total core flew. The factors, and axial power shapes. These effect of scram worth-scram delay time and rod factors are selected conservatively with insertion rate, all conservatively applied, are of respect to their effect on the applicable greatest significance in the early portion of the transient results as determined by the negative reactivity insertion. The rapid insertion current analysis model. This transient of negative reactivity is assured by the time require-model, evolved over many years, has been ments for the notch 46 (% 4%) and notch 38 (* 21%)
substantiated in operation as a conservative insertion.
The times for notch 24 (% 50%) and notch 04 (N 914) insertion are given to assure proper completion of the expected performance in the earlier portion a
of the transient, and to establish the ultimate fully shutdown steady-state condition.
Amendment No. 49 15
2.1 BASES (cont'd)
JAFEIPP APRM Flux Scram Trip Setting (Run Mode) (cont'd) d.
APRM Rod Block Trip Setting c.
rated power. This reduced flow referenced trip Reactor power level may be varied by moving control setpoint will result in an earlier scram during rods or by varying the recirculation flow rate. h slow thermal transients, such as the loss of 800F APRM system provides a control rod block to prevent feedwater heating event, than would result with rod withdrawal beyond a given point at constant re-the 120% fixed high neutron flux scram trip. De circulation flow rate, and thus provides an added g
i lower flow referenced scram setpoint therefore level of Protection before APRM Scram. This, rod 1
decreases the severity (A CPR) of a slow thermal block trip setting, which is automatically varieo j
transient and allows lower Operating Limits if with recirculation loop flow rate, prevents an in-l such a transient is the limiting abnormal crease in the reactor power level to excessive values due to control rod withdrawal. h flow variable operational transient during a certain exposure trip setting parallels that of the APRM Scram and interval in the cycle.
provides margin to scram, assuming a steady-state he APRM fixed high neutron flux signal does not operation at the trip setting, over the entire re-circulation flow range. h actual power distri-incorporate the time constant, but responds directly to instantaneous neutron flux. This bution in the core is established by specified scram setpoint scrams the reactor during fast power control rod sequences and is monitored continuously I
increase transients if credit is not taken for by the in-core LPRM system. As with the APRM scram I
a direct (position) scram, and also serves to trip setting, the APRM rod block trip setting is scram the reactor if credit is not taken for adjusted downward if the maximum fraction of limiting Power density exceeds the fraction of rated power, the flow referenced scram.
thus preserving the APRM rod block margin. As with The scram trip setting must be adjusted to ensure the scram setting, this may be accomplished by that the LHGR transient peak is not increased for adjusting the APRM gain.
any combination of maximum fraction of limiting I
power density (MFLPD) and reactor core thermal l
power. The scram setting is adjusted in accord-2.
Reactor Water Iow Imvel Scram Trip Setting (LLI) ance with the formula in Specification 2.1.A.I.c, when the MFLPD is greater than the fraction of h reactor low water level scram is set at a point rated power (FRP). This adjustment may be which will assure that the water level used in the Bases for the Safety Limit is maintained. The scram accomplished by either (1) reducing the APRM scram and rod block settings or (2) adjusting the setpoint is based on normal operating temperature indicated APRM signal to reflect the high peaking and pressure conditions because the level instru-sentation is density compensated.
condition, j
Analyses of the limiting transients show that no i
scram adjustment is required to assure that the l
MCPR will be greater than the Safety Limit when l
the transient is initiated f rom the MCPR operating limits provided in Specification 3.1.B.
i Amendment No. 49 la
JAFNPP 2.1 BASES (cont'd)
C.
References 1.
Linford, R. B., " Analytical Nethods of Plant Transient Evaluations for the General Electric Boiling Water Reactor",
NEDO-10802, Feb., 1973.
2.
" General Electric Fuel Application" NEDE 240ll-P-A (Approved revision number applicable at time that reload fuel analyses are performed).
20 Amendment No. 49 (Next page is 23) 1 2-
1.2 and 2.2 BASES JAFNPP The reactor coolant pressure boundary ANSI Code permits pressure transients up to integrity is an important barrier in the 20 percent over the design pressure (120s x prevention of uncontrolled release of 1,150 = 1,380 psig). The safety limit fission products. It is essential that pressure of 1,I75 psig is referenced to the the integrity of this boundary be pro-lowest elevation of the Reactor C olant o
tected by establishing a pressure limit System.
to be observed for all operating condi-tions and whenever there is irradiated The analysis in NEDO-24242, Supplemental fuel in the reactor vessel.
Reload Licensing Submittal f er James A.
FitzPatrick Nuclear Power PI it Reload 3, l
The pressure safety limit of 1,325 psig February 1980, shows that t main steam as measured by the vessel steam space isolation valve closure transient, with pressure indicator is equivalent to flux scram, is the most severe event re-l 1,375 psig at the lowest elevation of sulting directly in a reactor coolant the Reactor Coolant System. The 1,375 system pressure increase. The reactor psig value is derived from the design vessel pressure code limit of 1,375 psig, pressures of the reactor pressure given in FSAR Section 4.2, is above the vessel and reactor coolant system peak pressure produced by the event above.
piping. The respective design pressures Thus, the pressure safety limit (1,375 psig) are 1250 psig at 5750F for the reactor is well above the peak pressure that can vessel, 1148 psig at 56BoF for the re-result from reasonably expected overpressure circulation suction piping and 1274 psig transients. Figure 7 in NEDO-24242 presents 0
at 575 F for the discharge piping. The the curve produced by this analysis, pressure safety limit was chosen as the Reactor pressure is continuously indicated lower of the pressure transients permitted in the control room during operation.
by the applicable design codes: 1965 ASME Boiler and Pressure Vessel Code, Section A safety limit is applied to the Residual III for the pressure vessel and 1969 ANSI Heat Removal system (RHRS) when it is operating D31.1 code for the reactor coolant system in the shutdown cooling mode.
When operating piping. The ASME Boiler and Pressure in the shutdown cooling mode, the RHRS is I
vessel Code permits pressure transients included in the reactor coolant system.
up to 10 percent over design pressure (110% x 1,250 - 1,375 psig), and the Amendment No. 49 29
JAFNPP 4.1 SURVEILIANCE REQUIRENENTS 3.1 LIMITING COBOITIONS FOR OPERATION 4.1 REACTOR PROTECTION SYSTEM 3.1 REACTOR PR0ffECTION SYSTEM Applicability:
Applicability:
Applies to the instrismentation and associated Applies to the surveillance of the instru-mentation and associated devices which e
devices which initiate the reactor scram.
initiate reactor scram.
Objective:
Objective:
To assure the operability of the Reactor To specify the type of frequency of surveil-Protection System.
lance to be applied to the protection l
instrumentation.
Specification:
A.
% e setpoints, minimum number of trip systems, Specification:
minimum ntunber of instrasse.nt channels that must A.
Instrtamentation systems shall be be operable for each position of the reactor mode switch shall be as shown on Table 3.1-1.
functionally tested and calibrated as indli:sted in Tal les 4.1-1 and 4.1-2 The design system response time from the opening of the sensor contact to and including the respectively.
opening of the trip actuator contacts shall l
not exceed 50 usec.
B.
Minimum Critical Power Ratio (MCPR)
B.
Maxistan Fraction of Limiting Power Density (MPLPD)
During reactor power operation at rated power The MFLPD shall be determined daily during and flow, the MCPR operating limits shall reactor power operation at > 254 rated not be less than those shown belows thermal power and the APHM high flux scram and Rod Block trip settings adjusted if necessary as required by Specifications 2.1.A.1.c and 2.1.A.l.d, respectively.
l 9
l 30 Amendment No. 49 i
e
JAFNPP 3.1 (Cont'd)
FUEL MCPR OPERATING LIMIT FOR INCREMENTAL TYPE CYCLE 4 CORE AVERAGE EXPOSURE BOC4 to 2GWd/t EOC4-2GWd/t EOC4-1GWd/t before EOC4 to EOC4-1GWd/t to EOC4 f
At RBM trip Level Setting S = 0.66W + 39%
7x7 1.24 1.28 1.28 8x8 1.24 1.35 1.36 8x8R 1.24 1.35 1.36 C.
MCPR shall be determined daily during P8x8R 1.24 1.37 1.38 reactor power operation at > rated th mal power and following any change in pot i
At RBM Trip Level Setting S = 0.66W + 40 or 41%
level or distribution that would cause operation with a limiting control rod 7x7 1.27 1.28 1.2S pattern as described in the basec for Bx8 1.24 1.35 1.36 Specification 3.3.B.S.
8x8R 1.24 1.35 1.36 P8x8R 1.24 1.37 1.38 D.
When it is determined that a channel has failed in the unsafe condition, the At RBM Trip Level Setting S = 0.66W + 42%
other RPS channels that monitor the t
same variable shall be functionally 7x7 1.30 1.30 1.30 tested immediately before the trip 8x8 1.27 1.35 1.36 system containing the failure is tripped.
8x8R 1.25 1.35 1.36 The trip system containing the unsafe P8x8R 1.25 1.37 1.38 failure may be placed in the untripped t
condition during the period in which i
If anytime during reactor operation greater than surveillance testing is being performed 25% of rated power it is determined that the limiting on the other RPS channels.
value for MCPR is being exceeded, action shall then be initiated within fifteen (15) minutes to restore operation to within the prescribed limits. If the MCPR is not returned to within the prescribed limits within two (2) hours, an orderly reactor power re-duction shall be commenced immediately. The reactor power shall be reduced to less than 25% of rated i
power within the next four hours, or until the MCPR is returned to within the prescribed limits. For core flows other than rated, the MCPR operating limit shall be multiplied by the appropriate kg is as shown in figure 3.1.1.
Amendment No. 49 31 j
1
l JAFNPP
~.
4 h
3.1 BASES (cont'd)
Turbine control valves fast closure initiates a scram based on pressure switches sensing electro-hydraulic control (EHC) system oil pressure. The switches are located between fast closure solenoids and the disc l'
dump valves, and are set relative (500 < P < 850 psig) to the 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.
We requirement that the IRM's be inserted in the core when the APRM's read 2.5 b
indicated on the scale in the start-up i
and refuel modes assures that there is proper overlap in the neutron monitoring system fu.ctions and thus, that adequate coverage is provided for all ranges of reactor operation.
B.
The limiting transient which determines the required steady state MCPR limit depends on cycle exposure. The operating limit MCPR values as determined from the transient analysis for Cycle 4 (NEDO-24242) for various core exposures are given in Specification 3.1.B.
The ECCS performance analysis assumed reactor operation will be limited to MCPR, as described in NEDE-240ll-P-A. The Technical Specifications limit operation of the reactor to the more conservative MCPR based on consideration of the limiting transient as given in Specification 3.1.B.
Amendment No. 49 35
JAFNPP TABLE 3.1-1 (cont'd)
REACM PROTECTION SYSTEM (SCRAM) INSTidjMENTATION REQUIREMENT NOrrES OF TABLE 3.1-1 (cont'd)
C.
High Flux IRM D.
Scram Discharge Volume High Imvel E.
APRM 154 rower Trip l
7.
Not required to be operable when primary containment integrity is not required.
l 8.
Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
4 9.
1he APRM downscale trip is at.tomatically bypassed when the IRM Instrumentation is operable and not high.
10 An APRM will be considered oporable if there are at Icast 2 LPRM inputs per level and at least 11 LPRM inputs of the normal complement, i'
11.
See Section 2.1.A.I.
12.
This equation will be used in the event of operation with a maximum fraction of limiting power density (NFLPD) greater than the fraction of rated power (FRP).
where:
= Fraction of rated thermal power (2436 MWt)
MFLPD = Maximum fraction of limiting power density where the limiting power density is 18.5 KW/ft for 7x7 fuel and 13.4 MW/ft for 8x3, 8x8R and P8x8R fuel.
She ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used 6
W
= Icop Recirculation flow in' percent of rated (rated is 34.2 x 10 lb/hr)
Sn
= Scram setting in percent of initial 13.1he Average Power Range Monitor scram function 10 varied (Figure 1.1-1) as a function of recirculation loop flow (W).
She trip setting of this function must be maintained in accordance with Specification 2.1. A.I.c.
i Amendaient No. 49 43
JAFNPP 4
i 3.2 BASES (cont'd) crease to the Safety Limit. The trip logic The scaling arran9ement is such that trip for this function is 1 out of na e.g., any setting is less than a factor of 10 above trip on one of six APRM's, eight IRM's, or the indicated level.
four SRM's will result in a rod block.
The minimum instrument channel requirements is an indication the instrument has failed assure sufficient instrumentation to assure or the instrument is not sensitive enough.
the single failure criteria is met.
The In either case the instrument will not re-minimum instrument channel requirements for spond to changes in control rod motion and the RBM may be reduced by one for maintenance, thus, control rod motion is prevented. The testing, or calibration. This time period downscale trips are set at 2.5 indicated on is only three percent of the operating time scale.
In a month and does not significantly increase the risk of preventing an inadvertent The flow comparator and scram discharge control rod withdrawal, volume high level components have oaly one logic channgel and are not required for l'
The APRM provides gross core protections i.e.,
safety. The flow comparator must be by-j limits the gross core power increase from passed when operating with one recirculation i
withdrawal of control rods in the normal water pump.
l:
withdrawal sequence.
{~
The refueling interlocks also operate one The RBM rod block function provides local logic channel, and are required for safety protection of the cores i.e.,
the pre-only when the Mode Switch is in the Refuel-l vention of boiling transition in a local ing position.
I region of the core, for a single rod withdrawal error from a limiting control For effective emergency core cooling for i
rod pattern. The trips are set so that small pipe breaks, the HPCI system must MCPR is maintained greater than the safety function since reactor pressure does not t,
8 Limit.
decrease rapidly enough to allow either core spray or LPCI to operate in time.
[r The IRM rod block function provides local The Automatic pressure relief function i
as well as gross core protection.
is provided as a backup to the IIPCI in i
the event the HPCI does not operate.
i The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settin.p given in Amendment No. 49 58 i
l
\\
JAFNPP TABLE 3.2-3 INSTRUHFNTATION '111AT INITIATES CONTROL ROD BLDCKS Minista No.
Total Number of j
of Operable l
Instrument Instrument Trip Imvel Setting Instrument Channels Action I:
Provided by Design Channels Per for Both Channels i
Trip System
~
2 APRM Upscale (Flow Biased)
S<~ (0.66W+42t)x' FRP 6 Inst. Channels (1)
L)(FLPDj 2
APRM Upscale (Start-up i 12%
6 Inst. Channels (1)
'l Mode) 2 APRM Downscale 12.5 indicated on scale 6 Inst. Channels (1) 1 (6)
Rod Block Monitor S 1 0.66W+K (8) 2 Inst. Channels (1) l (Flow Biased) 1 (6)
Rod Block Monitor 12.5 indicated on scale 2 Inst. Channels (1)
I Downscale 3
IRM Downscale (2) 12% of full scale 8 Inst. Channels (1) l 3
IRH Detector not in (7) 8 Inst. Channels (1)
Start-up Position 3
IRM Upscale 186.4% of full scale 8 Inst. Channels (1) 2 (4)
SRM Detector not in (3) 4 Inst. Channels (1)
Start-up Position 2 (4) (5)
SRM Upscale 110 counts /sec 4 Inst. Channela (1) 5 NUrES FOR TABL" 3.2-3 1.
For the Start-up and Run positions of the Reactor Mode Selector Switch, there shall im two operable or tripped trip systems for each function. The SRM and IRM blocks need not be operable in run mode, and knendment No. 49 72 n
JAFNPP TABLE 3.2-3 (Cont'd)
INSTRUMENTATION TIIAT INITIATES CONTROL ROD HIDCKS NOTES FOR TABLE 3.2-3 (cont'd)
The APRM and RBH rod blocks need not be operable ils start-up mode. From and af ter the time it is found that the first. column cannot be se for one of the two trip systems, this condition may exist for up to seven days provided thac during that time the operable system is function: ally tested g
imanediately and daily tlwreaf ters if this condition lasts longer than seven days, tha system shall be tripped. From t:21 after the time it is found that the first column cannot be met for both trip systees, the mystems shall be tripped.
2.
IRM downscale is bypassed when it is on its lowest range.
3.
This function is bypassed when the count rate is > 100 cps.
4.
One of the four SRM inputs may be bypassed.
5.
This SRM Function is bypassed when the IRM range switches are on range 8 or above.
6.
%e trip la bypassed when the reactor power is < 30s.
7.
%is function is bypassed when the Mode Switch is placed in Run.
8.
S = Rod Block Monitor Setting in perc2nt of initial.
6 lb/hr).
W = Ioop recirculation flow in percent of rated (rated loop recirculation flow la 34.2 x 10 K = Intercept values of 394, 40%, 41%, and 42% can be used with appropriate MCPR Limits from Section 3.1.B.
Amendment No.
49 73
Q JAFNPP 3.3 (cont'd) 6.
During initial fuel loading or sub-6.
Prior to control rod withdrawal for sequent refueling, the restraints start-up or during refueling, verify imposed by Rod Sequence Control the conformance to Specificati w System groups A12 and A 4, B12 and 3.3.A.2.d before a rod may be by-3 B34 may be bypassed to perform the passed in the Rod Sequence Control required shutdown margin demon-System.
+
stration.
C.
Scram Insertion Times C.
Scram Insertion Timed 1.
The average scram insertion time, 1.
After each refueling outage all l
based on the de-energization of operable rods shall be scram time l
the scram pilot valve solenoids tested from the fully withdrawn as time zero, of all operable position with the nuclear system I
control rods in the reactor power pressure above 950 psig (with operation condit.lon shall be no saturation temperature). This greater thans testing shall be completed prior to exceeding 404 power.
Below Control Rod Average Scram 204 power, only rods in those Notch Position Insertion Time sequences (A12 and A34 or B12 and observed (Sec)
B34) which were fully withdrawn in the region from 1004 rod density 46 0.338 shall be scram time tested. During 30 0.923 all scram time testing below 204 24 1.992 power the Rui shall be operable.
04 3.554 l
r i
Amendment No.
49 95
i JAFNPP 3.3 (cont'd) 4.3 (cont'd) 2.
The average of the scram insertion 2.
At 8-week intervals, 15 percent of times for the three fastest operable the operable control rod drives shall control rods of all groups of four be scram timed above 950 psig. When-control rods in a two-by-two array ever such scram time measurements are shall be no greater thans made, an evaluation shall be made to provide reasonable assurance that Control Rod Average Scram proper control rod drive performance
,4 Notch Position Insertion Time is being. maintained.
observed (Sec) 46 0.361 38 0.977 24 2.112 l
04 3.764 4
Amendment No. 49 96
O JAFNPP 3.3 and 4.3 BASES (cont'd) rods have been withdrawn (e.g., groups A12 and
%1s system backs up the operator who A34), it is demonstrated that the Group Notch withdraws control rods according to made for the control drives is enforcea. This written sequences. The specified re-demonstration is made by performing the hardware strictions with one channel out of functional test sequence. The Group Notch re-service conservatively assure that straints are automatically removed above 204 power.
fuel damage will not occur due to rod withdrawal errors when this condition During reactor shutdown, similar surveillance
- exists, checks shall be made with regard to rod group availability as soon as automatic initiation of A limiting control rod pattern is a pattern the RSCS occurs and subsequently at appropriate which results in the core being on a thermal stages of the control rod insertion.
hydraulic limit (i.e., MCPR J.imits as shown in specification 3.1.B).
During use of a
'4.
The Source Range Monitor (SRM) System performs no such patterns, it is judood that testing automatic safety system functions i.e., it has no of the RBM System prior to withdrawal of scram function. It does provide the operator with such rods to assure its operability will a visual indication of neutron levol. The con-assure that improper withdraw does not sequences of reactivity accidents are functions of occur. It ac the responsibility of the the initial neutron flux. The requirement of at Reactor Analyst to identify these limit-I least 3 counts per sec assures that any transient, ing patterns and the designated rods either should it occur, begins at or above the initial when the patterns are initially established value of 10-8 of rated power used in the analyses or as they develop due to the occurrence of transient cold conditions. One operable SRM of inoperable control rods in other than channel would be adequate to monitor the approach limiting patterns. Other. personnel to criticality using homogeneous patterns of qualified to perform this function may sqattered control rod withdrawal. A minimum of be designated by the Plant Superintendent, two operable SRM's are provided as an added conservatism.
C.
Scram Insertion Times 4
5.
The Rod Block Monitor (RBM) is designed to auto-The Control Rod System is designed to bring matica11y prevent fuel damage in the event of the reactor subcritical at a rate fast erroneous rod withdrawal from locations of enough to prevent fuel damages i.e.,
to high power density during high power level prevent the MCPR from becoming less than operation. Two channels are provided, and the Safety Limit. Scram insertion time one of these may be bypassed from the console and scram reactivity curves shown in NEDO-for maintenance and/or testing. Tripping of 24242, Figures 2a, 2b and 2c were used one of the channels will block erroneous rod in analyses of power transients to determine i
withdrawal soon enough to prevent fuel damage.
MCPR limits. The scram insertion time test criteria of Section 3.3.C.1 conform to the scram insertion times of t*00-24242.
Therefore, the required protection is provided.
Amendment No. 49 102
l JAFNPP i
3.3 and 4.3 BASES (cont'd) later, control rod motion is estimated to actually begin. However, 200 maec is conservatively assumed for this time interval in the transient analysis and this is also included in the allowable scram insertion times of Specification
% e nianerical values assigned to the a;pecified 3.3.C.
We time to de-energize the pilot scram performance are based on the analysis of valve scram solenoid is measured during data from other BWR's with control rod drives the calibration tests required by Speci-the same as those on JAFNPP.
fication 4.1.
%e occurrence of scram times within the limits, ne scram times generated at each refuel-but significantly longer than the average, ing outage and during operation when com-should be viewed as an indication of a system-atic problem with control rod drives especially pared to scram times generated during pre-if the number of drives exhibiting such scram operational tests demonstrate that the times exceeds eight, the allowable number of control rod drive scram function has not deteriorated. In addition, each instant inoperable rods.
when control rods are scram timed during In the analytical treatment of the transients, operation or reactor trips, individual evaluations shall be performed to insure i
l 290 maec are allowed between a neutron sensor that control rod scram times have not reaching the scram point and the start of motion deteriorated.
of the control rods. This is adequate and con-j servative when compared to the typical time delay D.
Reactivity Anomalies of about 210 maec estimated from the scram test results. Approximately 90 maec of each of these During each fuel cycle, excess operative intervals result from the sensor and the circuit reactivity varies as fuel depletes and as delay, at this point, the pilot scram valve solenoid de-energizer. Approximately 120 msec any burnable poison in supplementary con-trol is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of 1
103 Amendiment No. 49 i
JAFNPp 3.4 and 4.4 BASES A.
Normal Operation poison peak. For a required pumping rate of 39 gal per min, the maximum The design objective of the Standby storage volume of the borcn solution Liquid Control System is to provide is established as 4,780 gal.
the capability of bringing the reactor from full power to a cold, menon-free Boron concentration, solution temper-shutdown assuming that none of the ature, and volume are checked on a withdrawn control rods can be inserted, frequency to assure a high reliability To meet this objective, the Standby of operation of the system should it Liquid Control System is designed to every be required. Experience with inject a quantity of boron which pump operability indicates that monthly produces a concentration of 600 ppm testing is adequate to detect if failures of boron in the reactor core in less have occurred.
than 125 min. Six hundred ppm baron concentration in the reactor core is The only practical time to test the required to bring the reactor from Standby Liquid control System is during full power to a subcritical condition a refueling outage and by initiation considering the hot to cold reactivity from local stations. Components of swing, decay of xenon poisoning, the system are checked periodically uncertainties and biases in the analyses, as described above and make a functional and an additional margin (25 percent) test of the entire system on a frequency for possible imperfect mixing of the of more than once each refueling outage l
chemical solution in the reactor water.
unne.assary. A test of explosive charges A minimum quantity of 2,500 gal. of from one manufs,+.uring batch is made solution having a 17 percent sodium to assure that L.~
-harges are satis-l pentaborate concentration is required factory. A continual check of the to meet this shutdown requirement.
firing circuit continuity is provided by pilot lights in the control room.
The time requirement (125 min) for insertion of the boron solution was The relief valves in the Standby Liquid selected to override the rate of Control System protect the system piping reactivity insertion due to cooldown and positive displacement pumps, which of the reactor following the xenon are nominally designed for 1,500 psig, Amendment No. 49 108
1 JAFNPP 3.5 (cont'd) 4.5 (cont'd) condition, that pump shall be considered 2.
Following any period where the LPCI inoperable for purposes satisfying Speci-subsystems or core spray subsystems fications 3.5.A, 3.5.C, and 3.5.E.
have not bcen required to be operable, the discharge piping of the inoperable system shall be vented from the high H.
Average Planar Linear Heat Generation Rata point prior to the return of the (APLilGR) system to service.
j l
'Ito APulGR for each type of fuel as a 3.
Wher. aver the HPCI, RCIC, or Core
,g fu.ction of average planar exposure shall Spray Syste.a is lined up to take not exceed the limiting value shown in suction from the condensate storage l
Figures 3.5.1 through 3.5.8.
If anytime tcnk, the discharge piping of the during reactor power operation greater HPCI, RCIC, and Core Spray shall than 25% of rated power it is determined be vented from the hig!: poJnt of that the limiting value for APLHGR is the system, and w.ater flow observed being exceeded, action shall then be on a monthly basis.
Initiated within 15 minutes to restore i
operation to within the prescribed limits.
4.
ne level switches located on the If the APulGR is not returned ta within Core ' ray and RHR System discharge the prescribed limits within two (2) hours, piping L, points which monitor an orderly reactor power reduction shall be these linea to insure they are full comenenced immediately. We reactor power shall be functionally tested each shall be reduced to less than 25% of rated month.
power within the next four hours, or until the APulGR is returned to within the pre-II. Aver.nge Planar Linear Heat Generation Rate scribed limits.
(APUIGR)
- l The APUlGR for each type of fuel as a function of average planar exposure shall be determined daily during reactor I
operation at > 25% rated thorinal power.
Amendment No. 49 123
JAFNPP 3.5 (cont'd) 4.5 (cont'd)
I.
Linear Heat Genration Rate (LHGR)
The linear heat generation rate (LHGR) of any I.
Linear Heat Generation Rate (LHGR) rod in any fuel assembly at any axial location shall not exceed the maximum allowable LHGR as The LHGR as a function of core height shall calculated by the following equations be checked daily during reactor operation 3
at > 25% rated thermal power.
~
LHCRmax, d. IJIGRd 1-AP/P),, (I/LT LHGRd = Design LHGR = G KW/ft.
i (AP/P) max = Maximum power spiking penalty = N LT = Total core length = 12 feet L = Axial position above bottom of core G = 18.5 KW/ft for 7x7 fuel bundles l
= 13.4 KW/f t for 8x8, 8x8R and P8x8R bundles N = 0.026 for 7x7 fuel bundles l
= 0.000 for 8x8, 8x8R and P8x8R fuel bundles If anytime during reactor power operation greater than 25% of rated power it is determined that *he limiting value for LHGR la being exceeded, act an shall then be initiated within 15 minutes to re-store operation to within the prescribed limits.
If the LiiGR la not returned to within the pre-acril>ed limits within two (2) hours, an orderly reactor power reduction shall be commenced imme-diately. The reactor power shall be reduced to less '.han 25% of rated power within the next four hours, or until the LHGR is returned to within the prescribed limits.
Amendment No.
49 124
JAFNPP 3.5 BASES A.
Core Spray 9ystem and Low Pressure of operable subsystems to assure the Coolant injection (LPCI) Mode of the availability of the minimum cooling RHR System systems. No single failure of ECCS equipment occurring during a loss-of-This specification assures that adequate coolant accident under these limiting emergency ccoling capability is available conditions of operation will result whenever irradiated fuel is in the reactor in inadequate cooling of the reactor vessel.
core.
Tht-loss-of-ooolant analysis is referenced core spray distribution has been shown, and described in General Electric Topical in full scale tests of systems similar Report NEDE-240ll-P-A.
in design to that of the FitzPatrick Plant, to exceed the minimum require-ments by at least 25 percent. In addi-tion, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel.
The accident analysis is additionally conservative in that no credit is taken The limiting conditions of operation for spray coolant entering the reactor in Specifications 3.5.A.1 through before the internal pressure has fallen 3.5.A 6 specify the combinations to 113 psig.
The LPCI mode of the RitR System is de-signed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system
)
is completely independent of the Core Spray Systems however, it does function i
g in combination with the core Spray System to prevent excessivs fuel clad temperature. The LPCI mode of 125 Amendment No. 49
JAFNPP 3.5 BASES (cont'd) requirements for the emergency diesel generators.
are within the 10 CFR 50 Appendix K limit.
The limiting value for APLHGR is shown in G.
Maintenance of Filled Discharge Pipe Figure 3.5.1 through 3. 5-8.
If the discharge piping of the core spray, LPCI, I. Linear Heat Generation Rate (LHGR)
RCIC, and HPCI are not filled, a water hammer can develop in this piping when the pump (s) are This specification assures that the linear started. To minimize damage to the discharge heat generation rate in any rod is less piping and to ensure added margin in the operation than the design linear heat generation.
of these systems, this technical specification 1
requires the discharge lines to be filled when-The LHGR as a function of core height shall ever the system is required to be operable. If be checked daily during reactor operation at a discharge pipe is not filled, the pumps that
>254 power to determine if fuel burnup, or supply that line must be assumed to be inoperable control rod movement has caused changes in for technical specification purposes. However, power distributulon. For LHGR to be a if. a water hammer were to occur, the system limiting value below 25% rated thermal power, would still perform its design function.
the ratio of local LHGR to average LHGR would have to be greater than 10 which is precluded H.
Average Planar Linear Heat Generation Rate (APLHGR) by a considerable margin when employing any permissible control rod pattern.
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CPh 50 Appendix K.
%e peak cladding temperature folloing a postu-lated loss-of-coolant accident is primarly a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly. Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than +200F relative to the peak temperature for a typical fuel design, the limit on the l
average linear heat generation rate is suf-I ficient to assure that calculated temperatures Amendment No. 49 130
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=
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- x,:
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5 10 15 20 25 30 9
PLANAR AVERAGE EXPOSURE (GWD/t)
FIGURE 3.5-7 fiAXI!1Uti AVERAGE PLANAR LINEAR HEAT GENERATION RATE (!!APLHGR) VERSUS PLANAR AVERAGE EXPOSURE RELOAD 3, P8DRB265L REFERENCE NEDO-24242 FULL CORE DRILLED SECTION 14 135e Amendment No. 49 m
14 3g 13 -
mb
<zW 5"
ms 12
=
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EE
=a BC 11 -
'M E *
!! =b ru<
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5 10 15 20 25 30 PLANAR AVERAGE EXPOSURE (GUL/t)
FICURE 3.5-9
!!AXI.'1U!! AVERAGE PLANAR LINEAR HEAT RATE (MAPLHGR) VERSUS PLAMAR AVERAGE EXPOSURE RELOAD 3, P8DRB283 REFERENCE NEDO-24242 FULL CORE DRILLED SECTION 14 135f Amendment No. 49
JAFNPP 5.0 DESIGN FEATURES B.
The reactor core contains 137 cruciform-shaped control rods as described in Section 3.4 of j
5.1 SITE the FSAR.
A.
The James A. FitzPetrick Nuclear Power Plant is located on the PASNY 5.3 REACTOR PRESSURE VESSEL portion of the Nine Mile Point site, approximately 3,000 ft. east of the The reactor pressure vessel is as Nine Mile Point Nuclear Station.
described in Table 4.2-1 and 4.2-2 The NMP-JAF site is on Lake Ontario of the FSAR.
The applicable design in Oswego County, New York, approxi-codes are described in Section 4.2 mately 7 miles northeast of Oswego.
of the FSAR.
{
The plant is located at coordinates narth 4,819, 545.012 m, east 386,968.945 m, 5.4 CONTAINMENT I
on the Universal Transverse Mercator System.
A.
The principal design parameters and characteristics for the B.
The nearest point on the property primary containment are given in line from the reactor building and Table 5.2-1 of the FSAR.
any points of potential gaseous effluents, with the exception of the B.
The secondary containment is as lake shoreline, is located at the described in Section 5.3 and the northeast corner of the property.
applicable codes are as. described This distance is approximately in Section 12.4 of the FSAR.
3,200 ft. and is the radius of the exclusion areas as defined in 10 CFR C.
Penetrations of the primary con-100.3.
tainment and piping passing through such penetrations are designed in accordance with standards set forth 5.2 REACTOR in Section 5.2 of the FSAR.
A.
The reactor core consists of not more than 560 fuel assemblies. For 5.5 FUEL STORAGE the current cycle four fuel types are present in the core:
- 7x7, A.
The new fuel storage facility is 8x8, 8 x 8R and P8 x 8R.
These fuel designed so that the K gg dry is e
types are described in Section 3.2 of 'he 0.90 and flooded is 0.95 des-FSAR and NEDO-24011.
The 7 x 7 fuel has 49 cribed in Section 9.2 of the FSAR.
fuel rods, the 8 x 8 fuel has 63 fuel rods and 1 water rod, and the 8 x BR and P8 x 8R fuel have 62 fuel rods and 2 water rods.
Amendment No. 3A, jpf 49 245
i 6.0 ADMINISTRATIVE CONTROLS Administrative controls are the means by which plant opera-tions are subject tr.. management control.
Measures specified in this section provide for the assignment of responsibilities, plant organization, staffing qualifications and related require-ments, review and audit mechanisms, procedural controls and reporting requirements.
Each of these measures are necessary to ensure safe and efficient facility operation.
6.1 RESPONSIBILITY The Resident Manager is recponsible for safe operation of the plant.
During periods when the Resident Manager is un-available, the Superintendent of Power will assume his responsibilities.
In the event both are unavailable, the Resident Manager may delegate this responsibility to other qualified supervisory personnel.
The Resident Manager reports directly to the General Manager and Chief Engineer for adminis-trative matters and functionally to the Manager - Nuclear Operations for operational related matters, as shown in Fig. 6.1-1.
6.2 PLANT STAFF ORGANIZATION i
The plant staff organization is shown graphically in Fig.
6.2-1 and functions as follows:
1.
A licensed senior reactor operator shall be on site at all times when there is fuel in the reactor.
2.
In addition to item 1 above, a licensed reactor operator shall be in the control room at all times when there is fuel in the reactor.
3.
In addition to items 1 & 2 above, a licensed reactor operator shall be readily available on site whenever the reactor is in other than cold condition.
4.
fwo licensed reactor operators shall be in the control room during startups and scheduled shutdowns.
j 5.
A licensed senior reactor operator shall be responsible for all movement of new and irradiated fuel within the site boundary.
A licensed reactor operator will be required to manipulate or directly supervise the mani-pulation of the controls of all fuel moving equipment, except the reactor building crane.
All fuel movements by the reactor building crane, except new fuel movements from receipt through dry storage, shall be under the direct supervision of a licensed reactor operator.
All
(
fuel movements within the core shall be directly monitored l
by a member of the reactor analyst group. (a)
Amendment No. h[ 49 247 SEE NEXT PAGE FOR FOOTNOTE l
)
i
- - ~ ~ _ _ _ _ _
Footnotes:
(a) Paragraph 5 is effective until the end of cycle 4.
During cycle 5 and thereafter, the following paragraph is effective:
5.
A licensed senior reactor operator shall be responsible for all movement of new and irradiated fuel within the site boundary. All fuel movement as defined by Technical Specifi-cation section 1.B., " Core Alterations," shall be directly supervised by either a licensed Senior Reactor Operator, or Senior Reactor Operator Limited to Fuel Handling, who has no other concurrent responsibilities during this operation.
J Amendment No. 49 247a g