ML20062L002
| ML20062L002 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/06/1981 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20062K996 | List: |
| References | |
| NUDOCS 8101120209 | |
| Download: ML20062L002 (18) | |
Text
' O ATTACHMENT I PROPOSED TECHNICAL SPECIFICATIONS CHANGES RELATED TO CORRECTION OF TYPOGRAPHICAL ERRORS AND EDITORIAL IMPROVEMENTS i
i POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333
~
JANUARY.6, 1981 8101120S07
JEr.TF 3.1 BASES The outputs of the subchannels are combined in a 1 out of 2 logic; i.e.,
an The reacto r protection system input signal on either one or both of automatically initiates a reactor scram the subchannels will cause a trip system to:
trip.
The outputs of the trip systems are arranged so that a trip on both 1.
Preserve the integrity of the systens is required to produce a reactor fuel cladding.
2.
Preserve the integrity of the This system meets the intent of IEEE -
279 (1971) for Nuclear Power Plant Protection Systems.
The system has a 3.
Minimize the energy which must reliability greater than that of a 2 out be absorbed following a loss of of 3 system and somewhat less than that coolant accident, and prevent of a 1 out of 2 system.
inadvertent criticality.
With the exception of the average power This specification provides the limiting range monitor (APRM)
- channels, the conditions for operation necessary to intermediate range monitor (Ifm) preserve the ability of the system to channels, the main steam isolation valve perform its intended function even closure and the turbine stop valve during periods when instrument channels
- closure, each subchannel has one may be out of service because of instrument channel.
When the minimum maintenance.
When necessary, one condition for operation on the number of channel may be made inoperable for brief operable instrument channels per j
intervals to conouct required functional untripped protection trip system is met tests and calibrations.
or if it cannot be met and the af f ected protection trip system is placed in a The Reactor Protection System is of the tripped condition, the effectiveness of dual channel type (Ref erence subsection the protection system is preserved.
7.2 FSAR).
The System is made up of two independent trip systems, each having Three APRM instrument channels are two subchannels of tripping devices.
provided for each protection trip Each subchannel has an input from at system-.
APRM's A and E operate contacts l least one instrument channel which in one subchannel and APRM's C and E monitors a critical parameter.
operate contacts in the other 32 Amendment No. 14 7
u
J
-n.
~
steam line isolation valves, main steam closure group. The water level instrumentation initiates protection for drain valves, recirc. sample valves (Group 1), initiates the IIPCI and RCIC the f ull spectrum of loss-of-coolant and trips the recirculation pumps. The accidents.
low-low-low reactor water level instru-Venturis are provided in the main steam mentation is set to trip when the water lines as a maans of measuring steam flow 3cvel is 18 in. above the top of the and also limiting the loss of mass active fuel. This trip activates the remainder of the CCCS subsystems, and inventory from the vessel during a stets line break accident. The primary.
starts the emergency diesel generators.
function of the instrumentation is to These trip level cettings were chosen to detect a break in the main steam line, be high enough to prevent spurious actu-For the worst case accident, main steam ation but low enough to initiate ECCS line break outside the drywell, a trip operation and prir.ary system 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 limiters and main steam line valve 10CFR100 will not be exceeded. For closure, limits the mass inventory loso large breaks up to the complete circumferential break of a 24 in.
such that fuel is not uncovered, fuel recirculation lino and with the trip temperature peak Lt approximately setting given above, ECCS initiation and 1,000"P and releaua of radioactivity to the environs is below 10CFh100 guide-
'primsry system isolation are initiated lines. Reference 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 B and 3 isolation valves. For the breaks Miscussed above, this instrumentation will generally initiate ECCS operation l
before the low-low-low water Icvel instrumentations thus the results given above are applicable here also.
See Specification 3.7 for isolation valve I
..lae'ndment No. 36, 4Af 56
SURVEILLANCE INSTRUMEffrATION Minimum No.
- of Operable No. of Channels Instrument Type Indication Provided Channels Instrument and Range by Design Action (Suppression Chamber Indicator
)
(Water Level Recorder
)
( (Wide Range)
-72 to +72 inches) 1
(
)
2 (2)
(Suppression Chamber Indicator
)
(Water Level Recorder
)
( (Harrow Range)
-6 to +6 inches )
N/A control Rod Indicator 1
(7)
Position Indication Postion 00 to 48 2
Source Range Indicator 4
(8)
Monitoru Recorder
[$[l 1 to 10 cps 6
3 Int.rmediate Indicator 8
(8) (9)
Range Monitor Recorder
-4 10 to 40% Rated Power 2
Average Power Indicator 6
(8) (9)
Recorder Range Monitor 0-125% Rated Power 1
Drywell-Suppression Recorder 2
(2)
Chamber Differential.
O to 5 pst Pressure Computer 0 to 5 psi Mart?.S FOR TABLE 3.2-6
- 1. From and af ter the date that the minimum number of operable instrument channels is one less than the minimum number specified for each parameter, continued operation is permissible during the succeeding 30 days unless the minimum number specified is made operable sooner.
- 2. In the event that all indications of this parameter is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condLtion in six (6) hours and a Cold Shutdown condition in the following eighteen (18) hours.
Amendment flo. Aff 76a
t..
J JADJPP TABLE 4.2-2 HI!!IMt!M TEST AND CALIBRATIO!! FRET'ID!CY FOR CORE At3D cot 1TAf tmENT COOLING SYSTEMS
. Instrument Channel Instrunent Punctional Test Calibration Frequency Instrument Check
- 1) kwactor Water level (1)
Once/3 months Once/ day
- 2) Drywell Pressure (1)
Once/3 months None t
- 3) Reactor Pressure (1)
Once/3 months None
- 4) Auto Sequencing Timers
!!A Once/ operating cycle None
(1)
Once/3 months None Pressure Interlock
- 6) Trip System Bus Power Monitors (1)
N/A None l
- 0) Core Spray Sparger d/p (1)
Once/ 3 months Once/ day 1
- 9) Steats Line High Flow (llPCI & RCIC)
(1)
Once/3 months None
- 10) Steam Line/ Area High Temp.(IIPCI & RCIC)
(1)
Once/ operating cyclo Once/ day
Once/3 months Hone
- 13) HPCI Suction Source levels (1)
Once/3 months None
- 14) 4KV Emergency Power Under-Voltage Relays Once/ operating cycle Once/ operating cycle Hone and tir.ers,
~1)
Once/3 monthe None liigh
.171 LPrI/ Cross Connect Valve Position once/ operating cycle NA NA Note:
See listing'of notes following Table 4.2-6 for the notes referred to herein.
.M 79 Amendment No. 1 i
1 3 cnd I.,3 BASES (cont'd)
JAFNPP At power levels below 20's of rated, rod drop accident consequences are abnormal control rod patterns could acceptabic.
Control rod pattern produce rod worths high enough to be of g nstratnts above 20*. of rated power are concern relative to the 280 calories per 2 rip sed by power distribution requirements ns de
.rechn {ical Speci fications.ined in pcet ion 3.5.3.5 of the gram drop limit.
In this range, the 11101 Power level and itSCS constrain the cont rol rod sequence.and patterns to those which
[#
"l "* * * * ? "l'"i t f the itsCS function involve only acceptable rod worths.
I.
In s age turbine pressure.
Ilecause the instrument has an instrument error of 2 2". o f f ul l powe r, the noninal The Itod her t h Hinimizcr and the Itod instrument setting is 22*. of rated power.
Sequences Cinitrol System provide Power IcVel for automat ic cutout of automatic supervision to assure that the Illet function is sensed by feedwater l
out-of-sequence control rods will not and steam flow and is set manually at be wi-hdrawn or inserted; i.e.,
it 30*. of rated power to be consistent with Iimit s operator deviance fron planned the ItSCS setting, withdrawal sequences.
They serve as a hackup to procedural control of Funct ional testing of the IMI prior control rod sequences which limit to the start of control rod withdrawal l
the maximal reactivity worth of at startup, and prior to attaining 20%
cont rol rods, in the event that the rated thermal power dtiring rod insertion Itod Wort h flinimi zer i s out of service, while shutting down, wi!! cnsure reliable when required, a second licensed operation and minini:e the prol. ability operator or other qualified technical of the rod drop ae. ident.
plant employee The RSCS can be functionally tested can manisally ful fill' the cont rol rod prior to cont rol rod withdrawal for pat tern conformance funct ions of this reactor startup.
Ily selecting, for system.
In this case, the RSCS is example, A12 and at tempt ing to withdraw, backed up by independent procedural by one notch, a rod or all rods in cont rol to assure conformance.
cach other group, it can be determined l
that the A12 group is exclusive, ny The functions of the R'0! and itSCS ypassing to full-out all A
- rods, gy makc *it unnecessary to specify a select,ng A34 and attempt,ng to withdraw, i
i y one n ch, a rgd or all rods in group licen'sc linit on rod worth to precludo the A34 group is determ_ned exclusive.
i unacceptable consequences in the event 1he same procedure can he repeated for of a cont rol rod drop. At low powers, below 20"., these devices force adherence the B groups.
A f.t er 50'. of the cont rol to accclitable rod patterns.
Above 20".
of rated power, no const raint on rod 10 1 pat tern i:. required to assure that Amendment No. 30
JAPHPP 3.] and 4.3 BASES (cont'd) rode have been withdrawn (e.g., groups A32 and h is system backs up the operator who A34), it is demonstrated that the Group Notch wittedraws control rods according to made for the control drives is enforced. %is written sequences. The specified re-demonstration is made by perforating the hardware strictions with one channel out. of functional test sequence. W e Group Notch re-service conservatively assure that
(
straints are automatically removed above 20s power.
f uel damage will not occur due to rod wittwirawal errors when this condition I
(
During reactor shutdown, similar surveillance exists.
checks shall be made with regard to rod group evallability as soon as automatio initiation of A limiting control rod pattern is a pattern the IGCS occurs and subsequently at appropriate which results in the core being on a thermal stages of the control rod insertion.
hydraulio limit (i.e., MCPR limits as shown in spooification 3.1.D).
During use of 4.
We Source Range Monitor (SitM) System performe no such patterns, it is judged that testing autossatto safety system functions i.e.,
it has no of the RDM System prior to withdrawal of scram function. It does provide the operator with such rode to assure its operability will a vistal indication of neutron level. %e con-assure that improper withdraw does not occur.
It is the responsibility of the sequences of reactivity accidents are functions of
- n the initial neutron flux. no requirement of at Reactor Analyst to,1dentify these limit-least 3 counts per suo assures that any transient, Ing patterns and the designated rods either abould it occur, begins at, or above the initial when the patterns are init.lally established value of 10 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 clu nnel would be adequate to monitor the approach limiting patterns. Other qualified to criticality using nosmogeneous patterne of personnel may perform this function.
sqattered control rod withdrawal. A minimwn of 4
two operable SitM's are provided as an added C.
Scram Insertion Times conservatism.
5.
We mod Block Monitor (ItBM) is designed to auto-We Control Rod System is designed to brjng satically prevent fuel damage in the event of the reactor ent> critical at a rate f ast erroneous rod witletrawal frium locations of enough to prevent fuel damages i.e.,
to l
high power density during high power level prevent the MCPR froen becoming less than operation.
'two channels are provided, and the Saf ety Limit..
Scram insertion time one of these may be bypassed frose the console and scram reactivity curvos shown in Ntmo-for malsitenance and/or testing. Trlpping of 24242, rigures 2a, 2h and 2c were used one of t.he channelu will block erroneous rod in analyses of power transients to deterimine withdrawal soon enough to prevent fuel damage.
HCPR limits. %e scram insertion time tout criteria of Section 3.3.C.1 conform to the scram insertion times of HIJM)-24242 Therefore, the required protection is provided.
Amendiment No. 47 102 4
i ATTACHMENT II SAFETY EVALUATION l
[
RELATED TO i
PROPOSED CHANGES TO THE TECHNICAL l
r SPECIFICATIONS INVOLVING CORRECTION OF TYPOGRAPHICAL ERRORS AND EDITORIAL IMPROVEMENTS POWER %iTHORITY OF THE STATE OF NEW YORK JAMES s.. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333
'UANUARY 6,
1981~
l t
l l
l l
Section I - Description of Modification Correction of typographical errors and implementation of editorial improvements in the Technical Specifications.
Section II - Purpose of Modification To make needed changes to the Technical Specifications which correct typographical errors and make editorial improvements.
Section III - Impact of the Change These modifications will not alter the conclusions reached in the FSAR and SER accident analysis.
Section IV - Implementation of the Modification The modification as proposed will not impact the ALARA or Fire Protection Program at JAF.
Section V - Conclusion The incorporation of these modifications:
a) will not increase the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; b) will not increase the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; and c) will not reduce the margin of safety as defined in the basis for any Technical Specification, and d) does not constitute an unreviewed safety question.
Sec:aion VI - References (a)
JAF FSAR (b)
JAF SER
l ATTACHMENT I PROPOSED TECHNICAL SPECIFICATIONS CHANGES RELATED TO CORRECTION OF TYPOGRAPHICAL ERRORS AND EDITORIAL IMPROVEMENTS l
1 l
i i
I POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333
~
JANUARY.6, 1981 i
1 F
4
JEtWF 3.1 BASES The outputs of the subchannels are combined in a 1 out of 2 logic; i.e.,
an The reacto r protection system input signal on either one or both of automatically initiates a reactor scram the subchannels will cause a trip system to:
trip.
The outputs of the trip systems are arranged so that a trip on both 1.
Preserve the integrity of the systens is required to produce a reactor fuel cladding.
2.
Preserve the integrity of the This system meets the intent of IEEE -
279 (1971) for Nuclear Power Plant Protection Systems.
The system has a 3.
Minimize the energy which must reliability greater than that of a 2 out be absorbed following a loss of of 3 system and somewhat less than that coolant accident, and prevent of a 1 out of 2 system.
inadvertent criticality.
With the exception of the average power This specification provides the limiting range monitor (APRM)
- channels, the conditions for operation necessary to intermediate range monitor (IPM) preserve the ability of the system to channels, the main steam isolation valve perform its intended function even closure and the turbine stop valve during periods when instrument channels
- closure, each subchannel has one may be out of service because of instrument channel.
When the minimum taaintenance.
When necessary, one condition for operation on the number of channel may be made inoperable for brief operable instrument channels per intervals to con 6uct required functional untripped protection trip system is met j
tests and calibrations.
or if it cannot be met and the aff ected I
protection trip system is placed in a The Reactor Protection System is of the trioped condition, the effectiveness of dual channel type (Ref erence subsection the protection system is preserved.
7.2 FSAR).
The System is made up of two independent trip systems, each having Three APRM instrument channels are two subchannels of tripping devices.
provided for each protection trip Each subchannel has an input from at system. APRM's A and E operate contacts l least one instrument channel which in one subchannel and APRM's C and E monitors a critical parameter.
operate contacts in the other 32 Amendment No. 14
e p
ny, steam line isolation valves, main stcam closure group. The water level instruinentation initiates protection fo'r drain valves, recire. sample valves (Group 1), initiates the llPCI and RCIC the full spectrum of loss-of-coolant and trips the recirculation pumps. The accidents.
Iow-low-low reactor water level instru-mentation is set to trip when the water venturis are provided in the main steam level is 18 in. above the top of the lines as a means of measuring steam flow active fuel. This trip activates the and also limiting the loss of mass remainder of the ECCS subsystems, and inventory from the vessel during a stets line break accident. The primary.
starts the emerger.cy diesel generators.
function of the instrumentation is to These trip level settings were chosen to detect a break in the main steam line.
- be high enough to prevent spurious actu-For the worst case accident, main steam ation but low enough to initiate ECCS line break outside the drywell, a trip operation and primary system isolation so that post-accident cooling can be ac-setting of 140 percent of rated steam comp 11shed and the guidelines of flow in conjunction with the flow limiters and main steam line valve 10CTR100 will not be exceeded. For closure, limits the mass inventory losa large breaks up to the complete circumferential break of a 24 in.
such that fuel is not uncovered, fuel recirculation line and with the trip temperature peak at approximately setting given above, ECCS initiation and 1,000 F an1 releaua of radioactivity to 0
the environs is below 10CFR100 guide-primary system isolation are initiated lines. Deference Section 14.6.5 FSAR.
in time to meet the above critoria.
Reference paragraph 6.5.3.1 FFAR.
The high drywell pressure instru-mentation is a diverse signal for mal-functions to the water level instru-mentation and in addition to initiating ECCS, it causes isolation of Groups B and 3 isolation valves. For the breaks discussed above, this instrumentation will generally initiate ECCS operation l
before the low-lou-low water Icvel instrumentation thus the results given above are applicable here also. See Specification 3.7 for isolation valve nlaendment No. 16 46 56
TABIE 3.2-6 SURVEILLANCE INSTRUMENTATION Minissa No.
- of operable No of Channels Instrument Type Indication Provided Channels Instrument and Range by Design Action (Suppression Chamber Indicator
)
(Water IAvel Recorder
)
( (Wide Range)
-72 to +72 inches) 1
(
)
2 (2)
(Suppression Chamber Indicator
)
(Water Level Recorder
)
{ (Narrow Range)
-6 to +6 inches )
W/A Control Rod Indicator 1
(7)
Position Indication Postion 00 to 48 2
Source Range Indicator 4
(8)
Monitors Recorder
_j 1 to 10 cp, 6
3 Intermediate Indicator 8
(8) (9)
Range Nonitor Recorder
~4
.J to 40% Rated Power 2
Average Power le41cator 6
(8) (9)
Gecorder Range Monitor 0-125% Rated Power 1
Drywell-Suppression Recorder 2
(2)
Chamber Differential O to 5 pst Pressure Computer 0 to 5 psi MCrrES FOR TABI2 3.2-6
- 1. From and after the date that the minimum number of operable instrument channels is one less than the minimum number specified for each parameter, continued operation is permissible during the succeeding 30 days unless the minimum number specified is made operable sooner.
- 2. In the event that all indications of this parameter is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown conditien in six U.) hours and a Cold Shutdown condition in the following eighteen (18) bours.
Amendment flo. Af 76a
3 p
)
JATHPP Tant.E 4. 2-a HitIIMt!M TEST AND CAI.IBRATION FRFOfif t!CY FOR CORE At3D cot!TAlt.HENT COOLItJG SYSTEMS
. Instriunent Channel Instrument Functional Test Calibration Frequency Instrument Check
- 1) Reactor Water level (1)
Once/3 months Once/ day 21 Drywell Pressure (1)
Once/3 months None
- 3) Reactor Pressure (1)
Once/3 months None
- 4) Auto Sequencing Timers t!A Once/ operating cycle Hone
,51 ADS - LPCI or CS Pump Disch.
(1)
Once/3 rnonths None Pressure Interlock
- 6) Trip System Bus Power Honitors (1)
N/A Hone l
- 0) Core Spray Sparger d/p (1)
Once/ 3 months once/ day t.
(1)
Once/3 months None
- 10) Steam Line/ Area liigh Temp. (IIPCI & RCIC)
(1)
Once/ operating cycle Once/ day
- 12) IIPCI & RCIC Steam Line 14w Pressure (1)
Once/3 months None
- 13) HPCI Suction Source tevels (1)
Once/3 months None
=
- 14) 4KV Energency Power Under-Voltage Relays Once/ operating cycle Once/ operating cycle None and tir.ers,
Once/3 months Hon'a flign
- 17) LPCI/ Cross Connect Valvo Position Once/ operating cycle
!!A NA Note: See listing'of notes following Table 4.2-6 for the notes ieferred to herein.
Amendment No. I 79 i
-m
3,3 cnd /. 3 PASES (cont'd)
JAFNPP l
At power levels below 20% of rated, rod drop accident consequences are abnormal control rod patterns could acceptable.
Control rod pattern produce rod worths high eneugh to be of constraints above 20*. of rated power are i
concern relative to the 280 calories per imp sed by power distribution requireme.nts as de ined in Section 3.5.3.5 of these Techn{ical Speci fications.
gram drop limit.
In this range, the R101 Power level and itSCS constrain the cont rol rod o au om c cutout of the ItSCS function sequence.and patterns to those which is nn by hnt stage turbine pressure.
involve only acceptable rod worths.
llecause the instrument has an instrument error of e 2". of fulI power, the noninal The Itod Sm t h Minimizer and the Itod instrument setting is 22*. of rated power.
Scipience. Control System provide Power level for automat ic cutout of automatic supervision to assure that the Ill01 function is sensed by feedwater out-of-sequence cont rol rods will not and steam flow and is set manually at he withdrawn or inserted; i.e.,
it 30". of rated power to be consi stent with limit s operator deviance frun planned the RSCS setting.
wi t hdrawal st spiences.
They serve as a backup to proccitural control of Functional testing of the ul01 prior control rod sequences which limit to the start of control rod withdrawal the maximal reactivity worth of at startup, and prior to attaining 20%
cont rol rods, in the event that the rated thermal power during rod insertion Itod Wort h flinimi zer i s out of service, while shutting down, will ensure reliable when required, a second licensed operation and mininize the probability operator or ot her ipiali fied technical of the rod drop accident.
plant employee
~
The !!SCS can be functionally tested can manually fulfill the control rod prior to cont rol rod withdrawal for pat tern conformance functions of this reactor startup.
11y select ing, for system.
In this case, the RSCS is example, A12 and at tempt ing to withdraw, backed up by independent procedural by one notch, a rod or all rods in cont rol to assure conformance.
cach other group, it can be determined that the A33 group is exclusive.
Ily The functions of the R101 and RSCS I)ypassjng to full-out all A
- rods, l2 make it unnecessary to specify a selecting A34 anil attempt,ng to withdraw, i
license linit on rod worth to preclude y one n tch, a rod or all rods in group the A34 group is determined exclusive.
unacceptable conseipiences in the event of a control rod drop.
At low powers,
.The same procedure can be repeated for below 20".,
these devices force adherence the 11 groups.
Af t er 50*. of the cont rol to acceptahic rod patterns.
Above 20".
of rated power, no const raint on rod 10 1 pattern is required to assure that Amendment Ko. 30
JAFNPP I
3.3 and 4.3 BASES (cont'd)
I rods have been withdrawn (e.g., groups A32 and
%1s system backs up the operator who A34), it is desnonstrated that the Group Hotch withdraws control rods according to made for the control drives is enforced. nis written sequences. The specified re-demonstration is made by performing the hardware strictions with one channel out of functional test sequence. % e Group Notch re-service conservatively assure that straints are autoanatically removed above 20s 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 availah!!!ty as soon as automatto initiation of A 11miting 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.
hydraulio limit (i.e., HCPn limits as shown in specificattori 3.1.D).
During use of
- 4 We Source Range Monitor (SRM) Systest perfome no such patterns, it is judged that testing automatic safety system functions i.e.,
it has no of the RnN System prior to wit:,L.swal of scram tunction.
It does provide the operator witli auch rods to asuure its operability will a visual indication of neutron level. %e con-assure that improper withdraw does not occur.
It is the responsibility of the sequences of reactivity accidente are functions of the initial neutron flux. % e requirement of at Reactor Analyst to identify these limit-a least 3 counts per suo assures that any transient, ing patterns and the designated rods either should it occur, begins at or above the initial when the patterns ase 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 SHH of inoperable control rods in other than channel would be ade<3uate to monitor the approach limiting patterns. Other qualified to criticality uutng homogeneous patterns of personnel may perform this function.
scattered control rod withdrawal. A minimius of two operable SHH's are provided as sa added C.
Scram Insertion Times conservatism.
5.
We mod alock Monitor (HmH) is designed to auto-
%e Control Rod System is designed to bring matloally prevent fuel damage in the event of the reactor subcritical at a r ate f ant erroneous rod withdrawal f rosa locations of enough to prevent fuel damages i.e.,
to high power density during high power level prevent the HCPR frosa Lecoming less than operation.
'tw channels are provided, and the Safety Limit.
Scram insertion time one of these may be bypassed from the consolo and scram reactivity curves shown in NL'DO-for maintenance and/or testing. Tripping of 24242, t*1gures 2a, 2h and 2c were used one of the channelu will block erroneous rod in analyses of power transients to detensine withdrawal soon enough to prevent fuel damage.
HCPR limits. %e scram insertion time tout criteria of Section 3.3.C.1 conform to the scram insertion times of NtJ)O-24242 Therefore, the required protection is provided.
Amendaunt No. K 102
,.--,.-,,.n-
,---v.-
,---n
9 e
O ATTACHMENT II SAFETY EVALUATION RELATED TO PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS INVOLVING CORRECTION OF TYPOGRAPHICAL ERRORS AND EDITORIAL IMPROVEMENTS f
POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DANUARY 6,
1981~
Section I - Description of Modification Correction of typographical errors and implementation of editorial improvements in the Technical Specifications.
r Section II - Purpose of Modification To make needed changes to the Technical Specifications which correct typographical errors and make editorial improvements.
Section III - Impact of the Change These modifications will not alter the conclusions reached in the FSAR and SER accident analysis.
l Section IV - Implementation of the Modification The modification as proposed will not impact the ALARA or Fire Protection Program at JAF.
Section V - Conclusion f
The incorporation of these modificat:lons:
a) will not increase the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; b) will not increase the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; and c) will not reduce the margin of safety as defined in the basis for any Technical Specification, and d) does not constitute an unreviewed safety question.
Section VI - References (a)
JAF FSAR (b)
JAF SER
.