ML20207S693
| ML20207S693 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/16/1987 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20207S671 | List: |
| References | |
| NUDOCS 8703200183 | |
| Download: ML20207S693 (15) | |
Text
F' I'
ATTACHMENT I ~ TO JPN-87-13 PROPOSED CHANGE TO THE TECHNICAL SPECIFICATIONS REGARDING PRESSURE-TEMPERATURE LIMITS (PTS-86-05)
NEW YORK POWER AUTHORITY J AMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59 8703200183 B70316 PDR ADOCK 05000333-p PDR,
JAFNPP TABLE OF CONTENTS (cont'd)-
t Page:
F.
Minimum Emergency Core Cooling System F.
.122 Availability G.
Maintenance.of Filled Discharge Pipe
'G.
122; H.
' Average Planar Linear Heat Generation H.
123 Rate (APLHGR)
I.. Linear Heat Generation Rate'(LHGR)
.I.-
124 J.
Thermal Hydraulic Stability J.
2124a SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 13.6 Reactor Coolant System 4.6 136 A. -Pressurization and Thermal Limits A. '
136 B.
Deleted C.
Coolant Chemistry C.
139
.D.
Coolant Leakage D.
141 E.
Safety and Safety / Relief' Valves E.
142a
'F.
_ Structural Integrity P.
144 G.
Jet Pumps G.
144 H.
DELETED 1.
Shock Suppressors (Snubbers)
I.
145b 3.7 Containment Systems 4.7 165 A.
165 B.
Standby Gas Treatment System B.
181 C.
184 D.
Primary Containment Isolation Valves D.
185-3.8 Miscellaneous Radioactive Material Sources 4.8 214 3.9 Auxiliary Electrical: Systems 4.9 215 A.
Normal and Reserve AC Power Systems A.
215 B.
Emergency AC Power System B.
216 C.
Diesel Fuel C.
218 D.
aiesel-Generator Operability D.
220 E.
Station Batteries E.
221 F.
LPCI MOV Independent Power Supplies F.
222a G.
Reactor Protection System Electrical G.
222c Protection Assemblies 3.10 Core Alterations 4.10 227 I
A.
Refueling Interlocks A.
227 B.
Core Monitoring B.
230 C.
Spent Fuel Storage Pool Water Level C.
231 D.
Control Rod and Control Rod Drive Maintenance D.
231 j
3.11 Additional Safety Related Plant Capabilities 4.11 237 A.
Main Control Room Ventilation A.
237 B.
Crescent Area Ventilation B.
239 C.
Battery Room Ventilation C.
239 Amendment No.
2, 4
11
~
r.
JAFNPP LIST OF FIGURES Figure Title Page 3.1-1 Manua'l Flow Control 47a 3.1-2 Operating Limit MCPR versus 47b
-4.1-1
_ Graphic Aid in the Selection of an Adequate Interval Between Tests 48 4.2-1 Test Interval vs. Probability of System Unavailability 87 3.4-1 Sodium Pentaborate Solution of System Volume-Concentration Requirements 110 3.4-2 Saturation Temperature of Sodium Pentaborate Solution 111 3.5-1 Thermal Power and Core Flow Limits of Specifications 3.5.J.1 and 3.5.J.2 134 3.5-6 (Deleted) 135d 3.5-7 (Deleted) 135e 3.5-8 (Deleted) 135f 3.5-9 MAPLHGR Versus Planar Average Exposure Reload 4 P8DRB284L 135g 3.5-10 MAPLHGR Versus Planar Average Exposure Reloads 4 & 5, P8DRB299 135h 3.5-11 MAPLHCR Versus Planar Average Exposure Reload 6, BP8DRB299 1351 3.6-1 Reactor Vessel Pressure-Temperature Limits 163 l 4.6-1 Chloride Stress Corrosion Test Results at 500*F 164
-6.1-1 Management Organization Chart 259 6.2-1 Plant Staff Organization 260 Amendment No.
7 vil
JAFEPP 3.6 LIMITING CONDITIONS FOR OPRRATION 4.6 SURVEILLANCE REOUIREMENTS 3.6 REACTOR COOLANT SYSTEM
.4.6 REACTOR COOLANT SYSTEM Applicability:
Applicability:
Applies to the operating status of the Reactor Coolant Applies to the periodic. examination and t'esting System.
requirements for the Reactor Coolant System.
l Objective:
Objective:
To assure the integrity and safe operation of. the To deterinine the condition. of the Reactor Coolant Reactor Coolant System.
System,and the operation of the safety devices related to it.
Specification:
Specification:
A.
Pressurization and Thermal Limit 5 A'.
Pressurization and Thermal Limits
~
1.
Reactor Vessel Head Stud Tensioning 1.
Reactor Vessel Head Stud Tensioning
. hen in the cold condition, the reactor The reactor vessel head bolting ' studs shall W
not be under tension unless the temperatures vessel head flange and the reactor vessel of the reactor vessel flange and the reactor flange temperatures shall be recorded:
j head flange are at least 90*F.
a.
Every 12 ' hours ' when the reactor vessel lI head: flange is(120*F and the studs arc tensioned.
b.
Every 30 minutes when the reactor vessel i
head flange is $100*F and the studs are tensioned.
l c.
Within 30 minutes prior to and every 30 minutes during tensioning _ of reactor
-vessel head bolting studs.
1 j
2.
In-Service Hydrostatic and Leak Tests 2.
In-Service Hydrostatic and Leak Tests During in-service hydrostatic or leak During' hydrostatic and. leak-testing the testing the Reactor Coolant System pressure Reactor-Coolant System pressure-and and temperature shall be on or to the right temperature shall be recorded every 30 of curve A shown in Figure. 3.6-l'. and the minutes. until 'two consecutive temperature maximum temperature change during any one readings are-within 5*F of each other.
hour period shall be:
l Amendment No. 14, 136 i
)
.JAFEPP
}
3.6 (cont'd) 4.6 (cont'd)-
a.
6,20*F when to the left of curve C.
l l
b.
n 100*F when on or to the right of j
curve C.
3.
Non-Nuclear Heatup and Cooldown 3.
Non-Nuclear Heatup and Cooldown I
During heatup by non-nuclear means During heatup by non-nuclear means, cooldown I
(mechanical),
cooldown following nuclear following nuclear: shutdown and low. power j
shutdown and low power physics tests the
. physics tests. -the reactor coolant system i
. pressure and pressure and temperature shall be recorded j
temperature shall be on or-to the right of every 30 minutes until two consecutive the curve B shown in Figure 3.6-1 and the temperature readings are within 5*F. of each maximum temperature change during any one other.
hour shall be $100*F.
4.
Core Critical Operation 4.
Core Critical Operation i
During all modes of operation with a
During all modes of operation with a
critical core (except for low power physics critical core (except-for low power physics tests) the reactor Coolant System pressure tests) the reactor Coolant System ' pressure -
and temperature shall be at or to the right and temperature shall be recorded within 30 1
of the curve C shown in Figure 3.6-1 and the minutes ~ prior to L withdrawal. of control rods maximum temperature change during any one to bring the reactor critical and every 30 l
hour shall bed 100*F.
minutes i
during heatup until two consecutive i
temperature readings are within 5'F of. each l
S.
With any of the limits of 3.6.A.1 through other.
3.6.A.4 above exceeded, either
\\
~
j a.
restore the temperature and/or pressure j
to within the limits within 30 minutes, j
perform an engineering evaluation to i
determine the effects of the out-of-j limit condition on the structural integrity of the. reactor coolant system, and determine that the reactor coolant i
system remains acceptable for continued l
operations; or Amendment No. h 137
JAFNPP b.
te in at least HOT SHUTDOWN within 12
'Y hours and ~ in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
6.
Idle Recirculation Loop Startup 6.
Idle Recirculation Loop Startup When Reactor Coolant System temperature is Within 30 minutes. Prior to startup of an
'7 140*F'an idle recirculation loop shall not idle loop:
be started unless:
a.
The temperature differential between the reactor coolant system and. the reactor reactor coolant system and the. reactor vessel bottom head drain line shall be vessel bottom head drain line is 6145*F,-
recorded, and and b.
When both loops are-
- idle, the b.
When both loops are
- idle, the differential temperature between the temperature difference between the reactor coolant system and the. idle loop reactor coolant system and the idle loop to be started shall be recorded, or to be started is 650*F, or c.
When only one loop is idle,-
the c.
When only one loop is
- idle, the temperature differential between the temperature difference between the idle idle loop and the operating loop shall loop and the operating loop is d50*F.
be recorded.
7.
Reactor Vessel Flux Monitoring The reactor vessel Flux Monitoring Surveillance Program complies with the intent of the
- May, 1983 revision to 10 CFR 50,.
Appendices G and H.
.The next flux monitoring surveillance ' capsule shall be removed after 15 effective full power years (EFPYs) and the test procedures and reporting requirements shall meet the requirements of ASTM E 185-82.
B.
Deleted B.
Deleted Amendment No.
138-
JAFNPP 3.6 and 4.6 BASES
(
region will not be warmed ' at ' an excessive rate
- A.
pressurization and Thermal Limits due to rapid sweep-out of cold coolant in the The reactor vessel design specification requires
-vessel lower head region by recirculation pump that the reactor vessel be designed for a maximum operation (cold coolant can accumulate as a heatup and cooldown rate of the contained fluid result of control rod drive inleakage and/or low (water) of 100*F/hr averaged over a period of 1 recirculation ficw rate during startup or hot hour.
This rate has.been chosen based on past I standby).
The limit on idle recirculation loop experience with operating power plants.
'The startup avoids high thermal stress effects in the associated time periods for heatup and cooldown pumps and piping, while also minimizing thermal cycles when the 100*F/hr rate is applied provide I stresses'on the vessel nozzles.
for efficient, but safe, plant operation.
The nil-ductility transition (NDT) temperature The reactor vessel manufacturer has designed the RTBDT is defined as the temperature below which vessel to the above temperature criterion.
In ferritic steel breaks in a brittle rather than the ccurse of completing the design, the manu-ductile manner.
Reactor vessel flux monitoring facturer performed detailed stress analysis.
samples are installed to conform with ' the 1972 i
This analysis includes more severe thermal con-draft revision of ASTM E 185.
Surveillance ditions than those which would be encountered program test results' have established the during normal heating and cooling operations, magnitude of changes in the NDT temperature as a function of the integrated neutron exposure for Specific analyses were made based on a heating BWR vessels.
The design life of the reactor and cooling rate of 100*F/hr applied continuously vessel is 40 years, and the maximum fast neutron s
over a temperature range of 100*F to 546*F.
exposure at 40 years was originally calculated to be 7.0 x
1017 n/cm.
Based on the 2
Calculated stresses were within 1965 ASME Boiler I surveillance program test
- results, the EOL i
and Pressure Vessel Code,Section III, with 1966 fluence is now estimated to be 1.7 x 1018 2
addenda stress intensity and fatigue limits.
The n/cm,
normal heating and cooling rate of 100*F/hr was also evaluated to assure protection against Fast neutron irradiation affects the fracture brittle fracture of the vessel shell remote from
. toughness of the reactor vessel material.
In discontinuities.
The rate meets the requirements order to assure that non-ductile failure does not of Appendix G to the Sununer 1972 Edition of 1971 occur, two types of information are needed*
l ASME
- III, throughout plant
- life, and is, therefore, satisfactory.
a) a relationship between the change in RTNDT and the accumulated fast neutron fluence, The limiting coolant temperature differential
- and, between the upper and lower regions of the reactor vessel, prior to recirculation pump b) a - relationship between the neutron fluence operation, assures that the vessel bottom head at the point of peak flux in the reactor pressure vessel shell and the effective full Amendment No.
power years.
146
.i c
.I
JAFNPP 3.6 and 4.6 BASES (cont'd)
The expected neutron fluence at the reactor
- junction, and one-quarter of the material vessel wall can be determined at any point during thickness at all other reactor vessel locations plant life based on the linear relationship.
and. discontinuity regions.
For the purpose of between the reactor thermal power output and the setting these operating limits, the reference s
corresponding number of neutrons produced.
temperature, RTEDT, of the' vessel material was '!'
Accordingly, neutron flux wires were removed from estimated from. impact. test data taken in the reactor vessel with the surveillance accordance with the requirements of the Code to specimens to establish the correlation at the which the vessel was designed and manufactured ;
capsule location by experimental methods.
The (1965 Edition including Winter 1966 addenda).
-flux distribution at the vessel. wall and 1/4 The RTNDT values for. the reactor vessel flange thickness (1/4T) depth
. was analytically region and for the reactor vessel shell beltline determined as a function of core height and region are 30*F, based on fabrication test :
{
azimuth to establish the peak flux location in reports.
The RTNDT for the remainder of the the vessel and the lead fac+or of the vessel is 40*F.
surveillance specimens.
The first surveillance capsule containing test :
A method of relating shift in RTNDT to specimens was withdrawn in April, 1985 after 6 accumulated fast neutron (>l MeV) fluence is EFPY.
The test specimens removed were tested contained in Regulatory Guide 1.99.
Experimental according to ASTM E 185-82 and the results are in 4
results of the evaluation of the irradiated GE report MDE-49-0386.
The curves of Figure surveillance specimens taken from the reactor 3.6-1, A through C,
reflect findings in the pressure vessel in April, 1985, show a shif t in report related to copper-phosphorus content of RTggy greater than predicted by Regulatory the reactor vessel shell beltline, flux wire Guide 1.99.
Therefore, the surveillance results testing fluence distribution analysis, and Charpy l were used with the methods of Regulatory Guide V-Notch specimen testing.
The next surveillance !
t 1.99 to establish the RTNDT shift.
The shift capsule will be removed after 15 EFPYs of for 16 EFPY was added to the unieradiated reactor operation and the results of the examination used pressure vessel shell beltline curve.
as a basis for revision of Figure 3.6-1 curves A, B and C for operation of the plant after 16 EFPYs. !
E Operating limits for the reactor vessel pressure and temperature during normal heatup and Figure 3.6-1 curve A establishes the minimum i cooldown, and during in-service hydrostatic and temperature for hydrostatic.and leak testing leak testing were established using 10 CFR 50 required by the ASME Boiler and Pressure Vessel Appendix G,
- May, 1983 and Appendix G of the Code,Section II.
Test pressures for in-service Sutnmer 1984 Addenda to Section III of the ASME hydrostatic and leak testing are a function of Boiler and Pressure Vessel Code.
These operating the testing temperature and the component limits assure that the vessel could safely material.
Accordingly, the maximum hydrostatic accommodate a postulated surface flaw having a test pressure will be 1.1 times the operating depth of 0.24 inch at the flange-to-vessel pressure or about 1105 psig.
Amendment No.
147
-- _. - - ~.. - - - - -.
.~
ma
.JAFEPP 1
3.6 and 4.6 BASgS (cont'd)
Figure 3.6-1, curve B, provides limitations for In the event of a steam line rupture outside the.
plant heatup and cooldown when the reactor is not drywell, with this coolant activity level, the critical or during low power physics _ tests.
The resultant radiological dose at the site boundary thennal limitation is based on -maximum heatup and
. would be 33 rem, to the thyroid - under adverse cooldown rates of 100*F/hr in any one-hour period.
meteorological. conditions assuming no more ' than l
3.ls(Ci/gm of dose equivalent I-131.
The reactor-l Figure 3.6-1, curve C,
establishes operating water sample will be used to assure that the l
limits when core -is critical.
These limits limit of Specification 3.6.C is not exceeded.
l include a margin of 40*F as required by 10 CFR 50 The total radioactive iodine activity would not i
Appendix G.
be expected to change rapidly over a period of 96 I
hr.
In _ addition,. the trend of the stack offgas The requirements for cold boltup of the reactor release rate, which is continuously monitored, is l
vessel closure are based on NDT temperature plus a good indicator of the trend of the iodine l
a 60*F factor of safety.
This factor.is based on activity in the reactor coolant.
Also during-i the requirements of the ASME Code to which the reactor startups and large power changes which i
I vessel was built.
For Figure 3.6-1,. curves _ A, B could affect iodine levels, samples of reactor and C,. 60*F margins are only added to the low coolant shall be analyzed to insure iodine j
temperature portion of the curve where concentrations are below allowable levels.
non-ductile failure is a concern.
The closure Analysis is required whenever the I-131 i
flanges have an NDT temperature not greater than concentration is within a factor of 100 of its l
30*F and are not subject to any appreciable allowable equilibrium value.
The necessity for j
neutron radiation exposure.
Therefore, the continued sampling following power and offgas minimum temperature of the flanges when the studs transients will be reviewed within 2 years of i
are in tension is 30*F plus 60*F, or 90*F.
initial plant startup.
i B.
Deleted The. surveillance requirements 4.6.C.1 may be satisfied by a
continuous monitoring system C.
Coolant chemistry capable-of determining the total iodine l
concentration in the coolant on a real time-j A radioactivity concentration limit of 20 gCi/ml
- basis, and annunciating at appropriate j
total iodine can be reached if the gaseous concentration levels such that sampling for j
effluents are near the limit as. set forth in isotopic ~ analysis can be initiated.
The design j
Radiological Effluent Technical Specification details of such a system must be submitted for 2.3.A if there is a failure or a prolonged l
evaluation and accepted by the Commission prior j
shutdown of the cleanup demineralizer.
to its implementation and incorporation in these Technical Specifications.
4 f
AmendmentNo,f 1
148 2
JAFNPP 1500-l VAllD TO 18 EFPY I
I
- - - - - -.l 1400 PSIG lA C
1400_
l 1400 PSIG g4ag pggy l ___ ___ _
190*F 221W 2619 ADJUSTED BELTLINE
~l
^l l
WITH 1/4T FLAW LMNTING l
,l
,I
.I I
- 5-. - ?"$ - - -
1200 - ~~-----*------.h----
i i
l l
l l
I 1
I I
I I
1000
- - ~ ~ ~ ~
l I
l I
l I
l l
1 i ----
N 800 l
O 755 PSIG l
', FW NOZZLE WITH 120V e
l l
I/
1/4T FLAW LIMITING b
l l
l l
N l
- - - d - - - -
j - - - - - - rl -- - - - - -
i 800 -
7--
D l
s y
FLANGE REGION l
l l
WITH 0.24 INCH l
.l
~
FLAW LIMITING W
I l
l l
400 - -.-------
I 312 PSIG l CURVE A -lNSERVICE NYDR0 STATIC &
l 90*F LEAK TESTING l CURVE B - NON-NUCLEAR NEATLMn &
I COOLDOWN 120*F l
CURVE C - CORE CRITICAL 200 -
r-----q-------
l 0,
57 P,SIG I
I I
0 I
i i
i i
i i
i i
i i
i i
50 73 100 125 150 175 200 225 250 275 300 325 350 TEMPERA TURE (* F)
FIGURE 3.6-1 REACTOR VESSEL PRESSURE - TEMPER ATURE LIMITS 183 Amendment No.
i ATTACHMENT 11 TO J PN 13 l-SAFETY EVALUATION FOR THE PROPOSED CHANGE TO THE TECHNICAL SPECIFICATIONS REGARDING PRESSURE-TEMPERATURE LIMITS (PTS-86-05) l l
i NEW YORK POWER AUTHORITY JAMES A.
FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59 l
1.- Description of the Proposed Change The proposed changes are limited to sections 3.6, 4.6, the related Table of Contents-pages, the related_ Bases, and Figure 3.6-1.
-The changes are:
Pace (s)
Section Chance ti.and
_. Table of These pages are corrected to account for vii Contents the changes made in this proposed amendment.
'136 3.6 and 4.6 Titles are realigned.
136-138 3.6.A and These sections are replaced with a new 4.'6.A description of limiting conditions and surveillance requirements for the reactor coolant system pressurization temperature.
This new description combined the current section 3.6.A/4.6.A (thermal limitations) and 3. 6. B / 4 '. 6. B (pressurization temperature).
This new description includes the following conditions: reactor vessel head stud tensioning, in-service hydrostatic and leak tests, non-nuclear heatup and cooldown, core critical operation, idle recirculation loop startup, and reactor vessel flux monitoring.
146-148 3.6 and This section is replaced with a new description 4.6.A of the bases used to establish operating Bases pressure and temperature limits for the reactor vessel.
146-148 3.6 and 4.6.B Bases This section is deleted.
148 3.6 and Change, " Radiological Effluent Technical 4.6.C Specification Section 3.2a" to " Radiological Bases Effluent Technical Specification 2.3.A."
163 Fig.
Replace with new figure, " Reactor Vessel 3.6-1 Pressure-Temperature Limits.
- 11. Purpose of the Proposed Changes I
In compliance with the requirements of 10 CFR 50 Appendix H, a
surveillance specimen removed from the FitzPatrick reactor was evaluated.
Based on this evaluation, new operating limit curves valid up to 16 effective full power years were developed.
The proposed i
E l
[
l k
g
- changes reflect the-new limits.
The proposed change also reflects ichanges in;the surveillance specimen withdrawal schedule.
bcr
.The purpose of the.new;11mits'is to assure reactor: vessel integrity.
.The limits accomplish this by restricting operating pressures and
. temperatures such that brittle fracture of the vessel cannot occur.
III. InDact of the ProDosed Chance
~
! Assuring reactor vessel integrity involves evaluation of the fracture toughness of the vessel ferritic materials.
The key values which characterize a, material's fracture toughness are the reference
- temperature of nil-ductility transition (RTNDT) and the upper shelf energy.
These are defined in 10 CFR 50 Appendix G and in the Appendix G of the ASME Boiler and Pressure Vessel Code,Section III.
These documents contain requirements used to establish the pressure-temperature operating limits-which.must be met to avoid brittle fracture.
The' requirements of Appendix G of 10 CFR 50 include safety margins for both critical and non-critical conditions.
Appendix H of 10 CFR'50 and ASTM-E185 establish the methods to be used for surveillance oof the' reactor vessel materials.
A method of relating shift in RTNDT to accumulated fast neutron fluence is described in Regulatory Guide 1.99.
In April, 1985 one of the surveillance specimen capsules required by Appendix H was removed from the James'A. FitzPatrick Nuclear Power Plant reactor and evaluated.-
Basedoon the results of this evaluation, operating limit curves' valid up to 16 effective full power years were developed for three reactor conditions:
hydrostatic pressure-testing, non-nuclear heatup and cooldown, and core critical operation.
These curves provide specific guidance for each mode of operation and are included in the
. proposed Technical Specifications to assure that operation in the brittle fracture range is avoided.
The proposed amendment to the Technical Specifications will have little impact on the' current: operation of FitzPatrick.
IV. Evaluation of Significant Hazards Considerations Operation of the FitzPatrick plant in accordance with the proposed amendment would not involve a significant hazard consideration as defined in 10 CFR 50.92 since it would not:
1.
involve a significant increase in the probability or consequences of an accident previously evaluated.
-Transient and accident analyses i
are based on reactor vessel integrity.
When these analyses were 1
originally done, operating limits were established to ensure that i
the temperatore and-pressure were kept in a safe range and reactor 1
vessel integrity would be ensured.
The proposed change will establish new,-more conservative limits based on new calculations and on the results of evaluation of surveillance specimens.
These l
limits were established according to the methods described in j
segulatory Guide 1.99 and 10 CFR 50 Appendix G, and incorporating the safety margins included in Appendix G.
Previous accident analyses are unaffected.
i i
e r
~. -
,mym.w, - y mmm y 3,
% %,, e
,,m.-p-m,,---.w.v,.
w--,,,,,.,v-,ver,-ww.,
_..-ww-.--,
- vr,,m,,,.w-ww,,..,w-
-s
~
2.
create.the possibility of a new or different kind of accident from
.any accident previously evaluated.
This change would only establish new pressure-temperature limits to ensure that' operation in the brittle fracture range is avoided.
The new limits were established according to NRC methodologies described'in 10 CFR 50 Appendix G.
The change restricts pressure at lower temperatures thus preventing operation in an unsafe region and could not create the possibility.
of a new type.of accident.
3.
involve a significant reduction in the margin of safety.
Accident analyses are based on reactor vessel integrity.
The analyses originally performed for the FitzPatrick plant established' operating limits which ensured that temperature and pressure were maintained in a safe range and vessel integrity was assured.
The proposed change will establish more conservative limits to ensure that the safety margin is maintained.
The proposed limits reflect neutron flux based on testing and calculations.
They incorporate safety margins as described in 10 CFR 50 Appendix G.
The effect of this change will be an overall improvement in plant safety, and assurance that the margin of safety is maintained.
In the April 6, 1983 Federal Register (48FR14870), the NRC published examples of license ~ amendments that are not likely to involve significant hazards considerations.
This proposed change most nearly corresponds to example vii, "A change to make a license conform to change in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations."
V.
Implementation of the Proposed Chance The proposed change will not adversely impact the ALARA. Security, or Fire Protection programs at the FitzPatrick plant, nor will it impact the environment.
VI. Conclusion The change, as proposed, does not constitute an unreviewed safety question as defined in 10 CFR 50.59, that is, it:
(a) will not increase the probability or the consequences of an accident or malfunction of equipment important to safety as evaluated previously in the safety analysis report:
(b) will not increase the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report:
(c) will not reduce the margin of safety as defined in the bases for a technical specifications:
(d) does not constitute an unreviewed safety question:
4
- (e)-involves no s'ignificant hazards-considerations. as defined in 10 CFR.50.92.
VII. References 1.
- James A.
FitzPatrick.FSAR Chapter 14.and'SER.
4 4
i i
a 4,
i l
l h
1
}.i-I, 4
I I
f l
_,,.__ _... _.