ML20002C206
| ML20002C206 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/06/1981 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20002C200 | List: |
| References | |
| NUDOCS 8101090618 | |
| Download: ML20002C206 (9) | |
Text
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1 l
i ATTACHMENT.I PROPOSED TECHNICAL SPECIFICATIONS CHANGES RELATED TO
(
l CORRECTION OF AND UPDATING OF MISCELLANEOUS TECHNICAL MATTERS l
\\
i POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 JANUARY 6,
1981 83o1 09 0 to8'
-e
a N
.1M 2.1 BASES (cont'd)
In order to ensure that the IRM provided adequate pro-high local peaks, and because several rods must be moved to change power by a significant percentage tection against the single rod withdrawal error, a
range of rod withdrawal accidents was analyzed. This of rated power, the rate of power rise is very analysis included starting the accident at various power slow. Generally, the heat flux is in near equili-levels. The most severe case involves an initial con-brium with the fission rate.
In an assumed uniform dition in which the reactor is just subcritical and rod withdrawal approach to the scram Icvel, the the IRM system is not yet on scale. This condition rate of power rise is no more than 5 percent of exists at quarter rod density. Additional conservatism rated power per minute, and the APRM system would was taken in this analysis by assuming that the IRM be more than adequate to assure a scram before channel closest to the withdrawn rod is by-passed, the power could exceed the safety limit. The 15 The results of this analysis show that the reactor is percent APRM scram remains active until the mode scrammed and peak power limited to one percent of switch is placed in the RUN position.
rated power, thus maintaining MCPR above the Safety Limit.
Based on the above analysis, the IRM provides protection against local control rod withdrawal errors
- c. APRM Flux Scram Trip Setting (Run Mode) and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
The APRM flux scram trip in the run mode consists of a flow referenced scram setpoint and a fixed b.
APRM Flux Scram Trip Setting (Re fuel or Startup and.
high neutron flux scram setpoint. The APRM flow referenced neutron flux signal is passed through a Hot Standby Modg) filtering network with a time constant which is For operation in the startup mode while the reactor is representative of the fuel dynmnics. This provides at low pressure, the APRM scram setting of 15 percent a flow referenced signal that approximates the of rated power provides adequate thermal margin between average heat flux or thermal power that is developed the setpoint and the safety limit, 25 percent of rated.
in the core during transient or steady-state condi-The margin is adequate to accommodate anticipated tions. This prevents spurious scrams, which have maneuvers associated with power plant startup.
Effects an adverse effect on reactor safety because of the of increasing pressure at zero or low void content are resulting thermal stresses.
Exampics of events minor, cold water from sources availabic during startup which can result in momentary neutron flux spikes 5
is not much colder than that already in the system, are momentary flow changes in the recirculation temperature coefficients are small, and control rod system flow, and small pressure disturbances 3
p nterns are constrained to be uniform by operating during turbine stop valve and turbine control procedures backed up by the rod worth minimizer and the valve testing. These flux spikes represent no Rod Sequence Control System. Worth of individual rods hazard to the fuel since they are only of a few is very low in a uniform rod pattern. Thus, of all seconds duration and less than 1207. of rated thermal possible sources of reactivity input, uniform control power.
rod withdrawal is the most probabic cause of signifi-cant power rise.
Because the flux distribution asso-The APRM flow referenced scram trip setting at full ciated with uniform rod withdrawals does not involve recirculation flow is adjustable up to 1177. of i
17 AmendmentNo.f,2[,f,/I
W f
- s. lZ s...
i' 1.0 (c:nd't)
JAFNPP I1%,
' '/
1.
Refuel Mode - The reactor is in main steam line isolation valve the refuel mode when the Mode closure trip is bypassed, the Switch is in the Refuel Mode Reactor Protection System-is
+
position.
When the Mode Switch energized with APEM (15 is in the Refuel position, the percent) and IR14 neutron monir refueling interlocks are ia toring system trins and control service.
rod withdrawal interlocks in service.
2.
Run Mode In this mode the reactor system pressure is at J.
Operable - A system or component l
or above 825 psig and the shall be. considered operable when it Reactor Protection System is is capable of performing its energized with APRM protection intended function in its required (excluding the 15 percent high rnanner.
flux trip) and the RBM i
interlocks in service.
K.
Ope ra t i ng - Operating means that a j
system or componont is performing l
3.
Shutdown Mode - The reactor is its intended functions in its I
in the shutdown mode when the required manner.
Reactor Mode Switch is in the Shutdown Mode position.
L.
Operatinc cycle - Interval between the end of one refueling outage and i
a.
Hot shutdown means condi-the end of the next subsequent tions as above with re-refueling outage.
actor coolant temperature
>2120 F.
M.
Primary Containment _ Integrity -
Primary containment integrity means b.
Cold shutdown means condi-ti st the drywell and pressure tions as above with re-suppression chamber are intact and actor coolant temperature all of the following conditions are i
52120 F and the reactor satisfied:
vessel vented.
1.
All manual containment isola-re.
Startup/ Hot Standby
- In this tion valves on lines connected mode the reactor protection to the Reactor Coolant System I
scram trips initiated by main or containment which are not j
steam line isolation valve required to be open during closure is bypassed when plant accident conditions are I ',
reactor pressure is less than closed.
These valves may be e
1,005
- psig, the low pressure I
Amendment No.
4
,.8
O e
clonure group.
The water level instrumentation initiaten protection fo'r
)
steam line isolation valves, main steam drain valves, cecirc. sample valves hhhhh (Group 1), initiates the llPCI and PCIC the full spectrum of loss-of-coolant Ej@j) and trips the recirculation pumps.
The accidents.
low-low-low reactor water 1cvel instru-Venturis are provided in the main steam
({ggg mentation is set to trip when the water lines as a means of measuring steam flow c
level is 10 in, above the top of the active fuel.
This trip activaten the and also limiting the loss of mans c
remainder of the ECCS subsystems, and inventory f r om the vessel during a stets g]{g line break accident. The pr i ma ry c===3 starts the emerger.cy diesel generators.
C]j{D function of the instrumentation is to These trip Icvel settings were chosen to detect a break in the main steam line.
cg, be high enough to prevent spurious actu-for the worst case accident, main steam b
ation but low enough to initiate ECCS line break cutside the drywell, a trip operation and prihary rystem isolation setting of 140 percent of rated steam so that post-accident cooling can be ac-flow in conjunction with.the flow complished and the guidelines of 1] miters and main steam line valvo 10CFR100 will not be er.ceeded.
For closure, limits the mass inventory loso large breaks up to the complete circumferential break of a 24 in.
nuch that fuel is not uncovered, fuct recirculation lino and with the trip temperature peak Lt approximately setting given above, ECCS initiation and 1,000"r and releaua of radioactivity to the environs is below 10CFR100 guide-primary system isolation are initiated lines. recference section 14.6.5 FSAR.
in time to meet the above criteria.
Reference paragraph 6.5.3.1 FSAR.
The high drywell pressure instru-mentation is a diverse signal for mal-functions to the water 1cvel instru-mentation and in addition to initiating ECCS, it causes isolation of Groups A and B' isolation valves. For the breaks Miscussed above, this instrumentation will generally initiate CSCS operation befo e the low-low-low water icvel inssi-nentations thus the results given above are applicable he e also, see Specification 3.7 for isolation valve
^-ene nn vt, n ?'
Ss
u 3.3 and 1. 3 BASES (cont'd)
JAFNPP s.
At power levels below 20% of rated, r d drop accident consequences are abnormal control rod patterns could acceptable. Control rod pattern produco rod worths high enough to be of constraints above 20t of rated power are concern relative to the 280 calories per impo nd by power distribution requirements gram drop limit, in this range, the Ill#f.
ns doftned in Section s3.3.3.5 of these and RSCS constrain the control rod
.I chnical Specifications.
Power level sequence and patterns to those which
[ r au matic cut ut the RES function involve only acceptable rod wortifs.
is unnd h & st stage tu dine pressure.
Because the instrument has an instrument The llod Worth Minimizer and the Rod error of s 2% of full power, the nominal instrument setting is 22% of rated power.
Sequence Control System provide Power Icvol for automatic cutout of automatic supervision to assure that out-of-sequence control rods will not the RWM function is sensed by g
steam flow and is set manually at he withdrawn or inserted; i.e., it limits operator devianco from planned 30% of rated power to be consistent with the itSCS setting, withdrawal sequences. They serve as a ba'kup to procedural control of c
Functional testing of the RWM prior control rod sequences which limit the maximal reactivity worth of to the start of control rod withdrawal at startup, and prior to attaining 20'.
control rods, in the event that the Itod Worth Minimizer is out of service, rated thermal power during rod insertion when rcquired, a second licensed while shutting down, will ensure reliable operation and minimizo the probability opercor or other qualified technical of the rod drop accident.
plant employee whose qualifications have been reviewed by the NPC The RSCS can be functionally tested can manually fulfill the control ' rod prior to control rod withdrawal for nattern conformance functions of this reactor startup.
By selecting, for
/ stem.
In this case, the RSCS is example, A12 and attempting to withdraw, backed up by independent procedural control to assure conformance.
by one notch, a rod or all rods in each other group, it,can be determined
, i The functions s,f the R!M and RSCS that the A12 group is exclusive.
By bypassing to full-out all A12 rods, make it unnecestsry to specify n selecting A,54 and attempting to withdraw, JItcense limi.t on rod worth to preclude by one notch, a rod or all rods in group unacceptable consequences in the event B, the A34 group is determined exclusive.
'of e control rod drop. At low powers, The same procedure can be repeated for below 204, these devices force adherence the H groups.
Af ter 50's of the control
',to acceptable rod patterns. Above 20%
of rated power, no constraint on rod pattern is. required to assure that 101 A m dent W. 9-
',6(coat,'d)
/
4 J;J},TP 3,6 (cont'd) too decirculation loop 5' have a flow imbalance of-1..
The 15 percent or more when the Q
name pu npa are operated at the
- peed.
of core indichted 'value M
The tlow rate varles from the value' 2.
' g::a flow' derived f roin loop c@::::::
ruea:airement a by noro than 10 gerectit.
3.
The diffuner to lower plenum readirot.
d'fferential preunure t
individual jet pumpt c,n an all f rom t.he ave r.uin of v.i r ies jet puicp dif f erent i;il prenaurert by niore th.in 10 piscerit.
.Tet Pu np riow Mic., atch i
1.
Po l lowi nej oue puinp op.2ra tion,
l
't.he siincharget valve of the J ott spee1 pnw.p inay riot he opened o
faster utile::su t he speed of the 1:. l e:in i han 50 perecut of
- nnp i t
- . rat e d c[ e ed.
2.
The reactor shall not be operated l
for a total period in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one or more recirculaLion loops out of service.
145 Amendment No. 14, 30
o.
s
[%
JMiiTP 3,6 and 4.6 BASr3 (cont'd) leakago path past would provide a the core thus reducing the core flow-'
The reverse flow through the rate.
inactive jet pump.would still be
{
iridicated by a positive dif ferential pressure but the net cffcct would be l
a slight decrease (3 percent to 6 percent) in the total core flow rne a sur ed.
This decrease, together with the loop flow increase, would result in a lack of correlation N
between measured and derived core flow rate.
- Finally, the affected jet pump diffuser diffcrential pres-sure signal would be reduced because the the backflow would be less than no rinal forward flow.
4, nozzle-riser system failure could A
failure o
also ger.erate the coincident body; however, the jet pump of a
converse is not true. The lack of any substantial stress in the jet pump body maken failure impossible without an initial nozzle-riser
~
system failure.
i
~
6 g
I e
h 5,
s 4
8
(
Amendment No.
I 155
- .., ~
s.
JAFHPP 3.6 and I. 6 BASES (cont'd) shutdown is required, the allowance of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to reach a cold shutdown condition will permit an orderly shutdown consistent Since with vrandard operating procedures.
plant startup should not commence with H.
Jet Pump Flow Mismatch knowingly defective safety related equipment, Specification 3.6.1.4 prohibits startup with inoperable snubbers.
Requiring the discharge valve of the lower speed loop to remain closed All safety related hydraulic snubbers are until the speed of the faster pump is below 50 percent of its rated visually inspected for ove. J 1 integrity and speed provides assurance when going operability. The inspection sTil include verification of proper orientation, adequate that from one to two pump operation
~
excessive vibration of the jet pump hydraulic fluid level and proper attachment risers will not occur, of snubber to piping and structures.
The inspection frequency is based upon maintaining Thus the I.
Ilydraulic Shock Suppressors a constant level of snubber protection.
required inspection interval varies inversely The number Snubbers are designed to prevent un-with the observed snubber failures.
restrained pipe motion under dynamic of inoperable snubbers found during a required loads as might occur during an earthquake inspection determines the time interval for the Inspections performed or severe transient, while allowing next required inspection.
f normal thermal motion during startup and before that interval has elapned may be used as The consequence of an a new reference point to determine the next shutdown.
the results of such early inoperable snubber is an increase in inspection. Ilowever,
the probability of structural damage to inspections performed before the original required piping as a result of a seismic or other initiating dynamic loads.
It is time interval has elapsed (nominal time less than may not be used to lengthen the required event 25s) inspection interval. Any inspection whose results therefore required that all snubbers required to protect the primary require a shorter inspection interval will override coolant system or any.other safety system the previous schedule, or component be operable during reactor operation.
Experience at operating facilities has shown that the required surveillance program should assure Because the snubber protection is required an acceptable Icvel of snubber performance low probability provided that the seal materials are compatible f
only during is allowed events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with the operating environment.
In case a for repairs or replacements.
156 5
4 Amendment No.
)RI
e 4A 3.5 (cont'd)
JAFNPP 4.5 (cont'd) 2.
From and after the date that one of 2.
When it is determined that one Core Spray the Core Spray Systems is nade or System is inoperable, the operable Core found inoperable for any reason, Spray System, the LPCI System, and the continued reactor operation is emergency diesel generators shall be permissible during the succeeding demcnstrated to be operable insnediately.
7 days unless the system is nade The remaining Core Spray System shall operable earlier, provided that be demonstrated to be operable daily during the 7 days all active com-thereaftar.
ponents of the other Core Spray System and the LPCI System and the emergency diesel generators shall be operable.
3.
The LPCI mode of the RHR System shall 3.
LPCI System testing shall be as be operable whenever irradiated fuel specified in 4.5.A.l.a, b, c, d, is in the reactor and prior to reactor f and g except that each RHR pump l
startup from a cold condition, except shall deliver at least 9,900 gpm j
as specified below.
against a system head corresponding to a reactor vessel pressure of a.
From the time that one of the 20 psig.
RHR pumps is made or found to be inoperable for any reason, con-a.
When it is determined that one tinued reactor operation is of the RHR pumps is inoperable, permissible during the succeeding the remaining active components 7 days unless the pump is made of the LPCI, containment spray operable earlier provided that subsystem, both Core Spray during such 7 days the remaining Systems, and the emergency active components of the LPCI, diesel generators required containment spray mode, all for operation shall be demon-active components of both Core strated to be operable immediately, Spray Systems, and the emergency and the remaining RHR pumps shall diesel generators are operable.
be demonstrated to be operable daily thereafter.
Amendment No. ZW 114