ML19284A876
| ML19284A876 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/03/1980 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML19284A874 | List: |
| References | |
| NUDOCS 8003110030 | |
| Download: ML19284A876 (25) | |
Text
ATTACliMENT A PROPOSED TECHNICAL SPECIFICATION CilANGES RELATED TO LICENSING OF RELOAD 3 AND OPERATION DURING CYCLE 4 8003310@D
, o., -
~
JAFNPP surveillance tests, checks, calibrations, and V.
Electrically Disarmed Control Rod 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.
In 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 commence at the end Electrical disarming does not eliminate of the original specified interval.
position indication.
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 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 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 (F LPD ).
b.
The testing of one system, subsystem, train or other designated component 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-pletely stable.
Amendment No.
6
JAFNPP 2.1 BASES 2.1 FUEL CLADDIE INTEGRITY The abnormal operational transients 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 of planned operating conditions up to the made by the model.
The comparisons and results thermal power condition of 2535 MWt.
The are summarized 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 maximum power level of Fitz-be about 25% 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 l e equivalent be exceeded.
to approximately 80% of the total scram worth of the control rods.
The 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 flow.
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 (N 4%) and notch 38 (% 21%)
substantiated in operation as a conservative insertion.
The times for notch 24 (% 50%) and notch 04 (% 91%)
insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.
Amendment No.
15
2.1 BASES (cont'd)
JAFNPP c.
APRM Flux Scram Trip Setting (Run Mode) (cont'd) d.
APRM Rod Block Trip Setting 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. The slow thermal transients, such as the loss of 8DoF 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 120s fixed high neutron flux scram trip.
The circulation flow rate, and thus provides an added lower flow referenced scram setpoint therefore level of protection before APRM Scram. This rod decreases the severity (A CPR) of a slow thermal block trip setting, which is automatically varied l
transient and allows lower Operating Limits if with recirculation loop flow rate, prevents an in-such a transient is the limiting abnormal crease in the reactor power level to excessive values i;
operational transient during a certain exposure due to control rod withdrawal.
The flow variable interval in the cycle.
trip setting parallels that of the APRM Scram and
[
provides margin to scram, asm* ing a steady-state The APRM fixed high neutron flux signal does not operation at the trip settb over the entire re-l incorporate the time constant, but responds circulation flow range.
actual power distri-directly to instantaneous neutron flux.
This bution in the core is occablished by specified scram setpoint scrams the reactor during fast power control rod sequences and is monitored continuously increase transients if credit is not taken for by the in-core LPRM system. As with the APRM scram 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 the flow referenced scram.
power density exceeds the fraction of rated power, 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 power density (MFLPD) and reactor core thermal power.
The scram setting is adjusted in accord-2.
Reactor Water Low Level Scram Trip Setting (LLI) ance with the formula in Specification 2.1. A.l.c, when the MFLPD is greater than the fraction of The 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 accomplished by either (1) reducing the APRM scram Bases for the Safety Limit is maintained.
The scram and rod block settings or (2) adjusting the setpoint is based on normal operating temperature indicated APRM signal to reflect the high peakf ng and pressure conditions because the level instru-condition.
mentation is density compensated.
Analysel of the limiting transients show that iso scram adfurtment is required to assure that the MCPR will be arcatcr than the Safety Ilmit when the transient is initiated from the MCPR operating limits provided in Specification 3.1.B.
Amendment No.
18
JAFNPP 2.1 BASES (cont'd)
~
C.
References 1.
Linford, R. B., " Analytical Methods 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.
(Next page is 23)
1.2 and 2.2 BASES JAFNPP The reactor coolant pressure boundary ANSI Code permits pressure transients ap 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 1,375 psig is referenced to the j
the integrity of this boundary be pro-lowest elevation of the Reactor Coolant 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 for James A.
FitzPatrick Nuclear Power Plant Reload 3, The pressure safety limit of 1,325 psig February 1980, shows that the 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-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 5680F for the re-result from reasonably expected overpressure circulation suction piping and 1274 psig transients. Figure 7 in NEDO-24242 presents at 5750F 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 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.
29
JAFNPP 3.1 LIMITING CONDITION 3 FOR OPERATION 4.1 SURVEILLANCE REQUIREMENTS 3.1 REACTOR PRCTTECTION SYSTDi 4.1 REACTOR PROTECTION SYST_EM Applicability:
Applicability:
Applies to the inst-umentation and associated Applies to the surveillance of the instru-devices which initiate the reactor scram.
mentation and associated devices which initiate reactor scram.
Objective:
Objective:
To assure the operability of the Reactor Protection System.
To specify the type of frequency of surveil-lance to be applied to the protection Specification:
instrumentation.
A.
The setpoints, minimum number of trip systems, Specification:
minimum number of instrument channels that must be operable for each position of the reactor A.
Instrumentation systems shall be mode switch shall be as shown on Table 3.1-1.
functionally tested and calibrated as The design system response time from the opening indicated in TM les 4.1-1 and 4.1-2 of the sensor contact to and including the re spectively.
opening of the trip actuator contacts shall l
not exceed 50 msec.
B.
Minimum Critical Power Ratio (MCPR)
B.
Maximum Fraction of Limiting Power Density (MFLPD)
During reactor power operation at rated power and flow, the MCPR operating limits shall The MFLPD shall be determined daily during not be less than those shown below:
reactor power operation at > 25% rated thermal power and the APRM high flux scram and Rod Block trip settings adjusted if necessary as required by Specifications 2.1.A.l.c and 2.1.A.I.d, respectively.
Amendment No.
30
JAFNPP 3.1 (Cont'd)
FUEL MCPR OPERATING LIMIT FOR INCRD1 ENTAL TYPE CYCLE 4 CORE AVERAGE EXPOSURE BOC4 to 2GWd/t EOC4-2GWd/t EOC4-lGWd/t before EOC4 to EOC4-lGWd/t to EOC4 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 thermal power and following any change in power 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 1.28 pattern as described in the bases for 8x8 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 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 condition during the period in which 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 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,>propriate kg is as shown in figure 3.1.1.
Amendment No.
31
JAFNPP 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 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 i
pressure.
The requirement that the IRM's be inserted in the core 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 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 Spccification 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.
35
JAFNPP TABLE 3.1-1 (cont'd)
~
REACTOR PROTECTICN SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT NOTES OF TABLE 3.1-1 (cont'd)
C.
High Flux IRM D.
Scram Discharge Volume High level E.
APRM 15% Power Trip 7.
Not required to be operable when primary containment integrity is not required.
O.
Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
g I
I 9.
The APRM downscale trip is automatically bypassed when the IRM Instrumentation is operable and not high.
10.
An APRM will be considered operable if there are at ' east 2 LPRM inputs per level and at least 11 LPRM inputs of the normal complement.
11.
See Section 2.1.A.1.
12.
This equation will be used in the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP).
where:
= Fraction of rated thermal power (2436 MWt) 5 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 8x8, 8x8R and P8x8R fuel.
The 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 W
= Icop Recirculation flow in percent of rated (rated is 34.2 x 106 lb/hr)
Sn
= Scram setting in percent of initial
- 13. The Average Power Range Monitor scram function is varied (Figure 1.1-1) as E. function of recirculation loop flow (W).
The trip setting of this function must be maintained in accordance with Specification 2.1. A.I.c.
Amer.inent No.
43
t JAFNPP 3.2 BASES (cont'd) crease to the Safety Limit. The trip logic The scaling arrangement is such that trip for this function is 1 out of n:
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.
A downscale indication on an APRM or IRM j
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 I
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 l
increase the risk of preventing an inadvertent The flow comparator and scram discharge control rod withdrawal.
volume high level components have only one i
logic channgel and are not required for l
The APRM provides gross core protection; i.e.,
safety. The flow comparator must be by-j limits the gross core power increase from passed when operating with one recirculation withdrawal of control rods in the normal water pump.
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 core:
1.e., the pre-only when the Mode Switch is in the Refuel-vention of boiling transition in a local ing position.
region of the core, for a single rod withdrawal error from a limiting control For effective emergency core cooling for 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 Limit.
decrease rapidly enough to allow either core spray or LPCI to operate in time.
I The IRM rod block function provides local The Automatic pressure relief function 5
as well as gross core protection.
is provided as a backup to the HPCI in the event the HPCI does not operate.
The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settirns given in Amendment No.
58
JAFNPP TABLE 3.2-3 INSTRUMENTATION THAT INITIATES CONTROL ROD BIDCKS Minimum No.
of Operable Total Number of Instrument Instrument Trip Invel Setting Instrt. ment Channels Action Channels Per Provided by Design Trip System for Both Channels
~
2 APRM Upscale (Flow Biased)
S<~ (0.66W+42%)x FRP 6 Inst. Channels (1) 82DJ 2
APRM Upscale (Start-up 1 12%
6 Inst. Channels (1)
Mode) 2 APRM Downscale
>2.5 indicated on scale 6 Inst. Channels (1) 1 (6)
Rod Block Monitor S 1 0.66W+K (8) 2 Inst. Channels (1)
(Flow Biased) 1 (6)
Rod Block Monitor
>2.5 indicated on scale 2..ast. Channels (1)
Downscale 3
IRM Downscale W
>2% of full scale 8 Inst. Channels (1) in (7) 8 Inst. Channels (1) 3 IRM Detector 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 5
2 (4) (5)
SRM Upscale 110 counts /sec 4 Inst. Channels (1)
NOTES FOR TABLE 3.2-3 1.
For the Start-up and Run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function. The SRM and IRM blocks need not be operable in run mode, and Amendment No.
72
JAFNPP TABLE 3.2-3 (Cont'd)
INSTRUMENTATION THAT INITIATES CONTROL ROD BIDCKS NOTES FOR TABLE 3.2-3 (cont'd)
The APRM and RBM rod blocks need not be operable in start-up mode. From and after the time it is found that the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped.
From and after the time it is found that the first column cannot be r.et for both trip systems, the systems 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 SFM inputs may be bypassed.
5.
This SRM Function is bypassed when the IRM range switches are on range 8 or above.
6.
The trip is bypassed when the reactor power is < 30%.
7.
This function is bypassed when the Mode Switch is placed in Run.
8.
S = Rod Block Monitor Setting in percent of initial.
W = Icop recirculation flow in percent of rated (rated loop recirculation flow is 34.2 x 106 lb/hr).
K = Intercept values of 39%, 40%, 41%, and 42% can be used with appropriate MCPR Limits from Section 3.1.B.
Amendment No.
73
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 cr during refueling, verify imposed by Rod Sequence Control the conformance to specification 12 and A 4, B12 and 3.3.A.2.d before a rod may be by-System groups A 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 Times 1.
The average scram insertion time, 1.
After each refueling outage all based on the de-energization of operable rods shall be scram time the scram pilot valve solenoids tested from the fully withdrawn as time zero, of all operable position with the nuclear system control rods in the reactor power pressure above 950 psig (with operation condition shall be no saturation temperature). This greater than:
testing shall be completed prior to exceeding 40% power.
Below Control Rod Average Scram 20% power, only rods in those Notch Position Insertion Time serpiences (Al2 and A34 or B12 and observed (Sec)
B34) which were fully withdrawn in the region from 100% rod density 46 0.338 shall be scram time tested. During 38 0.923 all scram time testing below 20%
24 1.992 power the RWM shall be operable.
04 3.554 Amendment No.
95
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 suen scram time measurements are shall be no greater than:
made, an evaluation shall be made to provide reasonable assurance that Control Rod Average Scram proper control rod drive performance Notch Position Insertion Tima is being maintained.
Observed (Sec) 46 0.361 38 0.977 24 2.112 04 3.764 Amendment No.
96
JAFNPp 3.3 and 4.3 BASES (cont'd) rods have been withdrawn (e.g., groups A12 and This system backs up the operator who A34), it is demonstrated that the Group Notch withdraws control rods according to made for the control drives it enforced.
This written sequences.
'1he 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 20% 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 limits as shown in specification 3.1.B).
During use of 4.
The Source Range Monitor (SRM) System performs no such patterns, it is judged that testing automatic safety system function; 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 level. The con-assure that improper withdraw does not sequences of reactivity accidents are functions of occur.
It is the responsibility of the the initial neutron flux.
The requirement of at Reactor Analyst to identify these limit-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 scattered 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 5.
The Rod Block Monitor (RBM) is designed to auto-The Control Rod System is designed to bring matically prevent fuel damage in the event of the reactor subcritical at a rate fast erroneous rod withdrawal from locations of enough to prevent fuel damage; 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. 7 tipping of 24242, Figures 2a, 2b and 2c were used one of the channels will block erroneous rod in analyses of power transients to determine 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 NEDO-24242.
Therefore, the required protection is provided.
Amendment No.
102
.TAFNPp 3.3 and ^ 3 BASES (cont' d) later, control rod motion is estimated to actually begin. However, 200 msec is conservatively assumed for this time interval in the transient analysis and this is also incivCed in the allowable The numerical values assigned to the specified scram insertion times of Specification scram performance are based on the analysis of 3.3.C.
The time to de-energize the pilot data from other BWR's with control rod drives valve scram solenoid is measured during the same as those on JAFNPP.
the calibration tests required by Speci-fication'4.1.
The occurrence of scram times within the limits, but significantly longer than the average, The scram times generated at each refuel-should be viewed as an indication of a system-ing outage and during operation when com-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 290 msec are allowed between a neutron sensor evaluations shall be performed to insure reaching the scram point and the start of motion that control rod scram times have not of the control rods.
This is adequate and con-deteriorated.
servative when compared to the typical time delay of about 210 msec estimated from the scram test D.
Reactivity Anomalies results. Approximately 90 msec of each of these intervals result from the sensor and the circuit During each fuel cycle, excess operative delay, at this point, the pilot scram valve reactivity varies as fuel depletes and as 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 103 Amendment No.
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 boron 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, xenon-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 systen 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 boron 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 chemical solution in the reactor water.
unnecessary. A test of explosive charges A minimum quantity of 2,500 gal, of from one manufacturing batch is made solution having a 17 percent sodium to assure that the charges are satis-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 pip _.g reactivity insertion due to cooldown and positive displacement pumps, which of the reactor following the xenon are nominally designed for 1,500 psig, r
Amendment No.
108
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 been required to be operable, the discharge piping of the inoperable system shall be vented from the high H.
Average Planar Linear Heat Generation Rato point prior to the return of the (APLHGR) system to service.
i The APLHGR for each type of fuel as a 3.
Whenever the HPCI, RCIC, or Core function of average planar exposure shall Spray System 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 tank, the discharge piping of the during reactor power operation greater HPCI, RCIC, and Core Spray shall than 25% of rsted power it is determined be vented from the high point of that the limiting value for APLHGR is the system, and water flow observed being exceeded, action shall then be on a monthly basis, initiated within 15 minutes to restore operation to within the prescribed limits.
4.
The level switches located on the If the APLHGR is not returned to within Core Spray and RHR System discharge the prescribed limits within two (2) hours, piping high points which monitor an orderly reactor power reduction shall be these lines to insure they are full commenced inmediately.
The reactor power sha]r be functionally tested each shall be reduced to less than 25% of rated mor ch.
power within the next four hours, or until the APLHGR is returned to within the pre-H. Average Planar Linear Heat Generation Rate scribed limits.
(APLHGR)
The APLHGR for each type of fuel as a function of average planar exposure shall be determined daily during reactor operation at > 25% rated thermal power.
Amendment No.
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 Ihe LHGR as a function of core height shall calculated by the following equation:
be checked daily during reactor operation g
at > 25% rated thermal power.
((AP/P) max (L/LT LIIGRmax, i LHGR
~
d 1-LIIGRd = Design LIIGR = G KW/ft.
(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/ft 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 the limiting value for LHGR is being exceeded, action shall then be initiated within 15 minutes to re-store operation to within the prescribed limits.
If the LIIGR is not returned to within the pre-scribed limits within two (2) hours, an orderly reactor power reduction shall be commenced imme-diately.
The reactor power shall be reduced to less than 25% of rated power within the next four hours, or until the LIIGR is returned to within the prescribed limits.
Amendment No.
124
JAFNPp 3.5 BASES A.
Core Spray System 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 cooling capability is available conditions of operation will result whenever irradiated fuel is in the reactor in inadequate cooling of the reactor vessel.
core.
The loss-of-coolant 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 RHR 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 System; however, it does function in combination with the Core Spray System to prevent excessive fuel clad temperature. The LPCI mode of 125 Amendment No.
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 requires the discharge lines to be filled when-The LHGR as a function of core height shall i
ever the system is required to be operable.
If be checked daily during reactor operation at I
a discharge pipe is not filled, the pumps that
>25% 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 l
if a water hammer were to occur, the system limiting value below 25% rated thermal power, l
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 dhat the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50 Appendix K.
The 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 average linear heat generation rate is suf-ficient to assure that calculated temperatures Amendment No.
130
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FIGU RE 3.5-7 11AXII1U?! AVERAGE PLANAR LINEAR HEAT GENERATION RATE (!!APLHGR) VERSUS PLANAR AVERAGE EXPOSURE RELOAD 3, P8DRB265L REFERENCE NEDO-24242 FULL CORE DRILLED SECTION 14 135e
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Plant Response to Generator Load Rejection, Without Bypass, EOC - 2000 mwd /t
l HEtthM-f sty 4
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Plant Response to Inadvertent Startup of HPCI Pump
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Plant Response to Feedwater Controller Failure, EOC4
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NEDO-24242 02 06 10 14 18 22 26 51 47 12 06 43 38 18 39 12 06 12 35 38 38 34 31 06 00 06 27 18 Notes:
1.
Rod Pattern Is 1/4 Core Mirror Synnetric, Upper Left Quadrant Shown on Map.
2.
Numbers Indicate Number of Notches Withdrawn out of 48.
Blank Is a Withdrawn Rod.
3.
Error Rod Is Rod (18,31).
Figure 6.
Limiting RWE Rod Pattern 18
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Figure 7.
Plant Response to MSIV Closurt
NEDO-24242 1.2 ULTIMATE PERFORMANCE LIMIT
= = = = = = = = = - = = = = - = = = = = = = = = = = = = = = = = = = = = = = = = = - - = = = =
1.0 -==========a-OR -
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NATURAL p
CIRCULATION y 0.6 5
8 0.4 106% ROD LINE 0.2 I
I I
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O 20 40 60 80 100 PERCENT POWER Figure 8.
Decay Ratio 20
NED0-24242 o
IN 1/ DELTA doyees C (MILLIONTHS)
~4
-8 z
W -12 9
i 8U a:
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8-16 O BOUNDING VALUE FOR 280 cal /g COLD D BOUNDING VALUE FOR 280 cal /g 286 C d CALCULATED VALUE -COLD h CALCULATED VALUE - 286 C
-24
_a i
I i
I l
I 400 800 1200 1000 2000 2400 FUEL TEMPER ATURE ( C)
Figure 9.
Doppler Reactivity Coefficient Comparisoa for RDA 21
NEDO-24242 20 16 O
E ii2 g
O m
l 5
t 208 5x O 800NO4NG VALUE FOR 200 cal /g O CALCULATED VALUE 4
i I
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at O
4 8
12 16 20 ROD POSITION (ft OUT)
Figure 10.
RDA Reactivity Shape Function at 20*C 22
NEDO-24242 20 O BOUNDING VALUE FOR 290 cal /g O CALCULATED VALUE IS O
E z 12 B
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Figure 11.
RDA Reactivity Shape Function at 286 C 23
NED0-24242 as 24 _
O SOUNDING VALUE FOR 200 cal /g O CALCULATED VALUE O
20 E
a i
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10 ELAPSED TIME (sec)
Figure 12.
RDA Scram Reactivity Function at 20*C 24
NEDO-24242 as O BOUNDING VALUE FOR 290 caWg O cAtcutAreD vAtuE 30 3
Cz<
h so 0
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N 5:p 15 N
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10 5
1 I
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to E LAPSED TIME (sec)
Figure 13.
RDA Scram Reactivity Function at 286*C 25/26
NED0-24242 APPENDIX A NEW BUNDLE LOADING ERROR EVENT ANALYSES PROCEDURE The bundle loading error analyses results are based on new analyses procedures for both the rotated bundle and the mislocated bundle loading error evcats.
The use of these new analyses procedures is discussed below.
NEW ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated bundle loading error event analysis results presented in this supplement are based on the new analysis procedure described and approved in Reference A-1.
This new method of performing the analysis is based on a more acct; ate detailed analytical model.
The principal difference between the previous analysis procedure and the new analysis procedure is the modeling of the water gap along the axial length of the bundle. The previous analysis used a uniform warer gap, whereas the new analysis utilizes a variable water gap which is more representative of the actual condition, since the interf acing between the top guide and the fuel spacer buttons, caused by miscrientation, causes the buidle to lean. The effect of the variable water gap is to reduce the power peaking and the R-factor in the upper regions of the limiting fuel rod. This results in the calculation of a reduced CPR for the rotated bundle. The calculation was performed using the same analytical models as were previously used. The only change is in the simulation of the water gap, which more accurately represents the actual geometry.
The results of the analysis indicate for the 8x8R bundle a 17.7 kW/ft LHGR and 0.17 ACPR (includes a 0.02 penalty due to variable water gap R-factor uncer-tainty) with a minimum CPR of 1.07.
NEW ANALYSIS PROCEDURE FOR THE MISLOCATED BUNDLE LOADING ERROR EVENT The mislocated bundle loading error event analyses results presented in this supplement are based on the new analysis procedures described in Reference A-1.
A-1
NED0-24242 This new method of performing the analysis employs a statistically corrected IIaling procedure and analyzes every bundle in the core.
The use of the statistically corrected llaling analyses procedure indicates that the LilGR is 15.7 kW/ft and that the minimum CPR for mislocated bundles is greater than the safety limit (1.07) for all exposures throughout Cycle 4.
REFERENCES A-1 Safety Evaluation Report (letter), D. G. Eisenhut (NRC) to R. E. Engel (GE), MFN-200-78, dated May 8, 1978.
A-2
NEDO-24242 APPENDIX B TRANSIENT ANALYSIS INPUTS Safety / relief valves Lowest Setpoint (psig) 1090 + 1%
Capacity at Setpoint (%)
86.7 B-1/B-2
NED0-24242 APPENDIX C GETAB TRANSIENT ANALYSIS INITIAL CONDITIONS Reactor Core Pressure (psia) 1035 Inlet Enthalpy (Btu /lb) 526.9 C-1/C-2
NEDO-24242 APPEhTIX D POWER SPIKING PENALTY 8x8, 8x8P., and P8x8R LHGR results reported for rod withdrawal error and loading error analyses include the 2.27 power spiking penalty for fuel densification.
D-1/D-2