ML20141B706
| ML20141B706 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/19/1997 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20141B703 | List: |
| References | |
| NUDOCS 9706240112 | |
| Download: ML20141B706 (24) | |
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4 Docket No. 50-423 S16514 Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Reactor Trio System Setooints (PTSCR 3-20-97)
Marked Uo Paae(s) l June 1997 ADOCK 05000423[
9706240112 970619 PDR p
U.S. Nuclear Regulatory Commission B16514\\ Attachment 2\\Page 1 MkRKUP OF PROPOSED REVISION Refer to the attached markup of the proposed revision to the Technical Specifications.
The attached markup reflects the currently issued version of the Technical Specifications listed below. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed markup.
Proposed Revision to Technical Specifications letter B15306, dated July 26, 1995, aNecting pages 2-9,2-10,2-11, and 2-12 is not reflected in the enclosed " Markup."
The following Technical Specification changes are included in the attached markup:
Tab'a Notations Note 1: Overtemperature AT is replaced with the attached.
Changes include: Certain function constants are deleted. Loop Specific indicated AT relocated to the left side of the algorithm. Constant definitions have been updated with minor wording changes and by changing "=" to "is" Numeric constants redefined using either "I or "#
l Table Notations Note 3: Overpower AT is replaced with the attached.
Changes include: Certain function constants are deleted. Loop Specific indicated AT relocated to the left side of the algorithm. Definitions have been updated with minor wording changes and by changing "=" to "is". Numeric constants redefined using either "I or "#
Table 2.2-1 Notes 1 and 3 Associated Bases changes Bases 2.2.1 - Overtemperature AT
!s
,d JABLE 2.2-1 (Continued)
- ~
~
'E TABLE NOTATIONS EDTE 1: OVERTEMPERA 7
I I
aT II + 'i ),
sai,(Kg-K s
4 (T
-T1+
(P - F ) - f (AI))
2 (1 + 7 SI gI+#3 g
E (1 + r S I+T5 5
6 3
U
[
i asured AT by React lant System Instrumentation; I
Whe AT I+rS g
- Lead-lag compens r on measured AT; f
- Time c tants uttitzed in Tag compensator for AT, rg 2 8 s.
2 s 3 s; l I+r5 N"#""
D 3
wrM ATtAttlEh 7
= TIM constants mM Ized in W 1ag compasatw f
,r3 - O s; j
3 e.
Indicated RATED THERMAL POWER; AT K
= 1.20 Four Loops Operating); 1.20 (
LoopsOperating);
g
.02456;
[
K
=
2I+#34
- The function generated the lead-lag compensator T,,, dynamic i
I I + #[
compensation; 3
{
7 i
- Time consta uttitzed in m lead-lag sator for T,,, r4 a 20 s, 4, r3
[
57 5 4 83
~_
i E
Ave ge temperature. *F; T
I compensator on mens
,9; c
3 r,
- Time constant utilized in the measured T,,, lag compensator, r6 ~ ' 53
[
l l
I3 l
TABLE 2.2-1 (Conttaued) t E
TABLE NOTATIONS (Continued) l t
NOTE 1: (Continued j
g T'
s 587.1*F (Nominal T atRATEDTHERMALPOKR);
[/E? LACE.
}
if M K
- 0.001311/pst-ArtgeM l
3 l
Press er pressure, psta; p
psta (Nominal RCS ope ingpressure);
P' i
Laplace transform tor, s ';
S t
and f 1s a function of the teated difference bet top and bottom detectors of the range neutron ten chambe ; wtth gains to be sel based on measured instrument response 5
ng plant startup tests s that:
j
- gb bet
-26% and + 3%, f,(AI) -,where g and gb ar# Percent RATED RMAL'P0E R l
(1) ForgItopand in th too halves of the core spectively,t and qt
- 9b is total 1 NR in percent of RAT TERMAL POKR; j
i (2) For each rcent that the magatt of g - g exceeds -26%, the AT T p Setpoint shall be l
automa ally reduced by 3.55%
itsvaImeatRAT[DTHERMALPOER*
i t
i For each percent that the stude of g exceeds +35, t AT Trip Setpoint shall be (3) automaticallyreducedby'l.981ofitsvakee-RAl[D TE POKR.
E h
a m
i
=
g NOTE 2:
The channel's maximum Trip Setpoint shall not exceed its competed Trip Setpoint by more than 1.4%
l!"
AT span (Four Loop Operatton); 2.7% AT span (Three Loop operation).
"q g
=
M i
{
i I
I
NOTE 1:
OVERTEMPERATURE AT
' AT' (1 +t s) s K -K (1 +t.s)(T-T')+KfP-P')-f (AI) i (AT,;(1+T s) 2 (1 +t s) 2 s
Where:
AT is measured Reactor Coolant System AT, *F; ATo in loop specific indicated AT at RATED THERMAL POWER, *F; (1 +t s) is the function generated by the lead-lag compensator on measured AT; i
(1+: 3) 2 t, and T2 are the time constants utilized in the lead lag compensator for AT, r, 2 8 sec. T2 s 3 sec; K, s 1.20 (Four Loops Operating); 51.20 (Three Loops Operating);
K 2 0.02456/ *F; 2
(1 + T.s)is the function generated Dy the lead-lag compensator for T4 (1 +r s) 5 t4 and tsare the time constants utilized in the lead-lag compensator for Tm,T4 2 20 sec, is s 4 sec; T is rneasured Reactor Coolant System average temperature. *F; T* is loop specific indicated Tavg at RATED THERMAL POWER, s 587.1*F; K 2 0.001311/ psi j
3 P is measured pressurizer pressure, psia; P' is nommal pressurizer pressure, 2 2250 psia;
[
s is the Laplace transform operator, sec ;
t I
i i
NOTE 1:
(Continued) and f,(AI) is a funchon of the indicated difference between top and bottom de'ectors of the power range neutron ion chambers; with nominal gains to be se%cted based on measured instrument response during plant startup tests calibrations such that For q - q. between -26% and + 3% f (AI) 2 0, where q and q. are percent RATED THERMAL POWER in the upper and lower (1) i halves of the core, respectively, and q + q. is the total THERMAL POWER in percent RATED THERMAL POWER; (2) For each percent that the magnitude of g - q exceeds -26%, the AT Trip Setpoint shall be automahcally reduced by 23.55% of its value at RATED THERMAL POWER; (3) For each percent that the magnitude of g - q exceeds +3% the AT Trip Setpoint shall be automatically reduced by 21.98% of its value at RATED THERMAL POWER.
b i
I I
I i
o b
r
~
TABLE 2.2-1 (Continued) 1 I"
TABLE NOTATIONS (Continued)
~
l y
g NOTE 3: DERPOER AT 7
II + T'SI
(
Ir I s AT, (K. - K,
! r,s
)
(
l I - T-] - f, (at))
[
(T - K, (T I I
AT E
(1 + r,5)
+ r,5)
(1 + r,5)
(1 + r,5)
(1 + r,5)
I w
Where:
AT As de ned in Note 1,
)
I + T'S M defined in Note 1, l
1 * '*I f27L ACE-U)iT+l As defined in te 1, MdM i
r, r,
/
I As defined in Note 1, j,,
7
)
[
A efined in Note 1, j
ATO As defined in Note 1, f
K, 1.09,
(
=
K,
' =
0.02/*F for 1 asing average t ature and 0 for decreasing erage j
temperatu f
'78 The tion generated by t rate-lag compensator for dynamic
(
1 + r,5 ompensation, a
=
i Time constants utti ed in the rate-lag c tor for T,y, r, it 10 s, l
r,
,o h
1 As defined in e1 y
I + r,5 "y
7,
- As deft in Note 1, b
i
.m If TABLE 2.2-1 (Continued)
TABLE NOTATIONS (Continuedl i
~%
Q NOTE (Continued) 0.00180/*F or T > T* and K6 - O for T T",
K t] g p J C E L
As ined in Note 1 F
T u.)T and K s 0/*F when T s T ; d s is the Laplace transform operator, see. l i i L March 11, 1991 ( LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Rance. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core i protection during reactor startup to mitigate the consequences of an uncon-i trolled rod clustcr control assembly bank withdrawal from a suberitical j condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes accive. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Trip System. Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors, and pressure is l within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically. varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the React tri is automat reduced according to the notations in Table 2.2-1. W secr ym@ Operation with a reactor coo an oop out of service requires Reactor Trip System modification. Three loop operation is permissible after resetting the K1 input to the Overtemperature AT channels, reducing the Power Range Neutron Flux High setpoint to a value just above the three loop maximum [ ) permissible power level, and resetting the P-8 setpoint to its three loop value. These modifications have been chosen so that, in three loop operation, each component of the Reactor Trip System performs its normal four loop function, prevents operation outside the safety limit curves, and prevents the DNBR from going below the design limit during normal operational and antici-pated transients. Overoower AT l I The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT MILLSTONE - UNIT 3 8 2-5 Amendment No. J2,60 0005 l t l \\ l Insert the following in the Bases for Overtemperature AT: "Although a direction of conservatism is identified for the l Overtemperature AT reactor trip function K and K gains, the gains 2 3 should be set as close as possible to the values contained in Note 1 to ensure that the Overtemperature AT setpoint is consistent with the assumptions of the safety analyses." I l l l l 1 { Docket No. 50-423 B16514 i j Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Reactor Trio System Setooints (PTSCR 3-20-97) Retvoed Paae(s) June 1997 l l U.S. Nuclear Regulatory Commission B16514\\ Attachment 3\\Page 1 RETYPE OF PROPOSED REVISION Refer to the attached retype of the proposed revision to the Technical Specifications. The attached retype reflects the currently issued version of the Technical Specifications. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications prior to issuance. Proposed Revision to Technical Specifications letter B15306, dated July 26, 1995, affecting pages 2-9,2-10,2-11, and 2-12 is not reflected in the enclosed " Retype." l i I 1 TABLE 2.2-1 (Continued) TABLE NOTATIONS DE 5; NOTE 1: OVERTEMPERATURE AT G -i E f ar ' (1+t s) (1+t,s) (T - T') +K (P -P') -f (AI) i 3 3 AT (1+t 8) (1+T 8) y oj 2 s E A Where: AT is measured Reactor Coolant System AT, *F; ATo is loop specific indicated AT at RATED THERMAL POWER, *F; (1+t s) i (1+t 8) is the function generated by the lead-lag compensator on measured AT; 2 r, and 72 are the time constants utilized in the lead-lag compensator for AT, r 1 8 sec, 72 1 3 sec; i K, s 1.20 (Four Loops Operating); i 1.20 (Three Loops Operating); toa K 2 0.02456/*F; 2 (1+t,s) (1+t s) is the function generated by the lead-lag compensator for T,; 3 7 and is are the time constants utilized in the lead-lag compensator for T, 74 2 20 sec, is s 4 sec; g 4 { T is measured Reactor Coolant System average temperature, 'F; g T' is loop specific indicated Tavg at RATED THERMAL POWER, s 587.l*F; K3 2 0.001311/ psi z P P is measured pressurizer pressure, psia; y P' is nominal pressurizer pressure,12250 psia; s is the Laplace transform operator, sec-'; "w gge TABLE 2.2-1 (Continued) TABLE NOTATIONS (Continued) o EE NOTE 1: (Continued) gg and f,(AI) is a function of the indicated difference between top and bottom detectors of the power range neutron ion chambers; with nominal gains to be selected based on measured instrument response during plant startup tests calibrations such that: (1) For q, - q, between -26% and +3%, f,,(AI) 10, where q, and q, are percent RATED THERMAL POWER in the upper and lower halves of the core, respectively, and q, + q, is the total THERMAL POWER in percent RATED THERMAL POWER; (2) For each percent that the magnitude of q, - a6 exceeds -26%, the AT Trip Setpoint shall be automatically reduced by 13.S5% of its value at RATED THERMAL POWER. (3) For each percent that the magnitude of q, - q, exceeds +3%, the AT Trip Setpoint shall be automatically reduced by 11.98% of its value at RATED THERMAL POWER. 93 o NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.4% AT span (Four Loop Operation); 2.7% AT span (Three Loop Operation). E
- =
CL Ea 5 M. _m t TABLE 2.2-1 (Continued) . t TABLE NOTATIONS (Continued) E5 i o r-t 5 NOTE 3: OVERPOWER AT E m ' aT ' (1+t s) (r.,s) i ( AT, (1+t 8) (1+Ti8) T -K, (T-7") s: K. -K3 g o 2 i Where: AT is measured Reactor Coolant System AT, *F; [ ATo is loop specific indicated AT at RATED THERMAL POWER, *F; i (1+t s) i (1+t 8) is the function generated by the lead-lag compensator on measured AT; 2 r, and 7 are the time constants utilized in the lead-lag compensator for AT, r,18 sec, 72 1 3 sec; f 2 K, s 1.09; K 10.02/*F for increasing Tavg and K, s 0 for decreasing T,; (s.,s) l ( 1 + t.,s) is the function generated by the rate-lag compensator for Tavg; k ry is the time constant utilized in the rate-lag compensator for Tavg, ry 110 sec;
- (
T is measured average Reactor Coolant System temperature, *F; E T" is loop specific indicated Tavg at RATED THERMAL POWER, s 587.l*F; I K.10.00180/*F when T > T" and K, s 0/*F when T 1 T"; s is the Laplace transform operator, sec". l O i ? $5 TABLE 2.2-1 (Continued)
- FM E
TABLE NOTATIONS (Continued) m E i G w NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.7% AT span. (Four Loop Operation) NOTE 5: Setpoint is for increasing power. l NOTE 6: Setpoint is for decreasing power. i ro I i l l a E a .I = ? LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Ranae. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the ensequences of an uncon-trolled rod cluster control assembly bank withdraA 1 from a subcritical condition. These trips provide redundant protection to 6 Low Setpoint trip of the Power Range, Neutron Flux channels. The Source age channels will initiate a Reactor trip at about 10' counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximably 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Trip System. Overtemoerature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power t distribution, provided that the transient is slow with respect to piping j transit delays from the core to the temperature detectors, and pressure is l within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by.the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the l I notations in Table 2.2-1. Although a direction of conservatism is identified for the Overtemperature AT reactor trip function K2 and K3 gains, the gains should be set as close as possible to the values contained in Note 1 to ensure that the l Overtemperature AT setpoint is consistent with the assumptions of the safety analyses. l Operation with a reactor coolant loop out of service requires Reactor l Trip System modification. Three loop operation is permissible after resetting l the K1 input to the Overtemperature AT channels, reducing the Power Range Neutron Flux High setpoint to a value just above the three loop maximum permissible power level, and resetting the P-8 setpoint to its three loop value. These modifications have been chosen so that, in three loop operation, each component of the Reactor Trip System performs its normal four loop function, prevents operation outside the safety limit curves, and prevents the DNBR from going below the design limit during normal operational and antici-pated transients. l Overoower AT l The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT MILLSTONE - UNIT 3 9 2-5 Amendment No. #, pp, 0630 L . =. i l Docket No. 50-423 B16514 l l Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Reactor Trio System Setooints (PTSCR 3-20-97) Backaround and Safety Assessment t i June 1997 i U.S. Nuclerr Regul: tory Commission B16514\\ Attachment 4\\Page 1 Nackaround Notes 1 and 3 to Technical Specifications Table 2.2-1 define the values for the i constants used in the Overtemperature Delta T (OTAT) and Overpower De!ta T (OPAT) Reactor Trip System Instrumentation setpoint calculators. Many of the constants are defined in a manner implying that the terms must be set exactly to defined values, it is not possible to achieve compliance because some tolerance must be allowed for calibration accuracy. Millstone Unit No. 3 instrument and Control surveillances allow a tolerance around the Technical Specification values for individual terms Ki, K2, K3, K4, K., K., T', T", P' and f (AI). These tolerances are equivalent to the equipment accuracy specifications and are included in the Channel Statistical Allowance provided by WCAP 10991, Rev. 4. Although nominal treatment is allowed for individual constants, plant surveillance procedures require the as left channel setpoints to trip at less than or equal to the setpoint defined by the OTAT and OPAT algorithms provided in Table 2.2-1. Safety Assessment This Proposed Technical Specification Revision defines the numeric constants using either "s" or "2:" The polarity of the operators indicate the direction of conservatism. In order to comply with this change in notation, minor changes are required to the procedures that calibrate the OTAT and OPAT functions. The acceptance criteria for components implementing the constants will be selected such that one limit of error will be equivalent to the value specified in the proposed Technical Specification change and the other limit will be in the direction of conservatism identified by the change. The direction of conservatism for the various terms can be determined by considering the purpose of the OTAT and OPAT trip functions, and by reviewing the algorithms. The OTAT trip protects the core from damage due to departure from nucleate boiling (DNB). DNB is associated with unbalanced power distributions, high temperatures and/or low pressures. OPAT ensures that the limit for allowable heat generation (kw/ft) is not exceeded. The OPAT setpoint is reduced when Tavg increases above nominal to account for temperature induced changes in density and heat capacity of water. The trip setpoints are continuously calculated based on fixed constants and process inputs for real time measurements of temperature, pressure and core power distribution. Additional changes have been made to the notes to reflect hardware modifications that were made during construction of Millstone Unit No. 3. These modifications eliminated hardware that was not required because the function constants were defined as zero in the existing notes to Table 2-2.1. The hardware change eliminated the filters on AT, I I (1 + r3s), Tavg, (1 + r,s), and f2(AI). i l U.S. Nuclear Regulatory Commission B16514\\ Attachment 4\\Page 2 l The proposed change will move the ATo term from the right side of the OTAT and OPAT l algorithms to the left side. This accurately reflects plant design, where AT power is l represented as the fraction of Rated Thermal Power (RTP) in the Process Protection System, AT/ATo. This is simply a change in notation and the algorithm is mathematically (although not functionally) equivalent to the original notation. This aspect of the change does not impact the margin of safety. The remaining definitions have been updated, changing "=" to "is." This is an administrative change and does not impact safety. A change to the bases is included per the recommendation of Westinghouse. The change adds a statement recommending that K2 and K3 be set as close as possible to the values provided in Note 1 to be consistent with the assumptions of the safety analyses. The proposed change is safe and does not create an unreviewed safety question, This evaluation demonstrates that this change does not increase the probability or consequences of a malfunction of the OTAT and OPAT channels. This assessment concludes that the margin of safety for transients identified in the FSAR is not reduced when the constants are adjusted in the directions identified by the proposed change. 1 I - l t l l l \\ l Docket No. 50-423 l B16514 I l l \\ Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Reactor Trio System Setooints (PTSCR 3-20-97) Sianificant Hazards Consideration and Environmental Considerations I i i June 1997 l U.S. Nuclecr Regulatory Comrnission l B16514\\ Attachment 5\\Page 1 l Sianificant Hazards Consideration NNECO has reviewed the proposed revision in accordance with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve an SHC because the revision would 1 not: 1. Involve a significant increase in the probability or consequence of an accident previously evaluated. l [ The proposed changes to Technical Specification Table 2.2-1 Notes 1 and 3 for i the addition of the inequalities ensure that the constants used for OTAT and OPAT will be set conservatively with respect to the assumptions in the accident analysis. The effect on the turbine runback function has been evaluated with respect to the Loss Of External Electrical Load And/Or Turbine Trip analysis and l it has been determined that this change does not increase the probability of this transient. The change was also reviewed to determine if it produced an increase in the probability of an unnecessary or spurious reactor trip and it was determined that it did not. This change does not increase the probability of any previously evaluated accident. The consequences of previously evaluated accidents, including Uncontrolled Rod Cluster Assembly Bank Withdrawal At Power, Rod Cluster Control Assembly Misalignment, Uncontrolled Boron Dilution, Loss Of External Electrical Load And/Or Turbine Trip, Excessive Heat Removal Due To Feedwater System Malfunctions, Excessive Load increase incident, Accidental Depressurization Of The Reactor Coolant System, Accidental Depressurization Of The Main Steam System, Loss Of Reactor Coolant From Small Ruptured Pipes Or From Cracks in Large Pipes Which Actuate ECCS, or Major Secondary System Pipe Ruptures have not changed. The administrative changes have no impact on the design or operation of Millstone Unit 3. Therefore, the proposed revision does not involve a significant increase in the i probability or consequence of an accident previously evaluated. 2. Create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes to Technical Specification Table 2.2-1 Notes 1 and 3 do l not alter the design, construction, operation, maintenance or method of testing of equipment. The proposed changes alter the Technical Specification description l of a OTAT and OPAT setpoint functions and requires only slight changes to the l actual setpoints in the field. The OTAT and OPAT functions serve to mitigate the i U.S. Nucisar R:gulatory Commission B16514\\ Attachment 5\\Page 2 i effects of accidents by opening the Reactor Tr p breakers or reduce power by " running back" turbine electrical load. The chlange does not create any j-interfaces to plant control or protection systems and therefore, no new mechanism for accident initiation has been introduced. The proposed change does not introduce the possibility of an accident of a different type than i previously evaluated. l Therefore, the proposed revision does not create the possibility of a new or l different kind of accident from any accident previously evaluated. 3. Involve a significant reduction in a margin of safety. The proposed changes to Technical S% ' -ation Table 2.2-1 Notes 1 and 3 do i not affect the integrity of any physical f.o.on protective boundaries, increase the delays in actuation of safety systems beyond that assumed in the safety analysis l or reduce th9 margin of safety of any system. These changes ensure that i actuation of Overtemperature AT and Overpower AT reactor trips will occur conservatively with respect to the assumptions of the accident analysis. ( Therefore, the proposed revision doce not involve a significant reduction in a margin of safety. l In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC. Environmental Considerations j i NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed revision does not involve j an SHC, does not significantly increase the type and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupations! radiation exposures. Based on the foregoing, NNECO concludes that the proposed revision meets the criteria delineated in 10CFRS1.22(c)(9) for categorical exclusion from the requirenents of an environmental considerations. .