View source for B17190, Application for Amend to License DPR-65,changing Tech Specs 2.1.1, Safety Limits - Reactor Core, 2.2.1, Limiting Safety Sys Settings - Reactor Trip Setpoints & 3.3.1.1, Instrumentation - Reactor Protective Instrumentation

{{Adams
| number = ML20236T263
| issue date = 07/21/1998
| title = Application for Amend to License DPR-65,changing Tech Specs 2.1.1, Safety Limits - Reactor Core, 2.2.1, Limiting Safety Sys Settings - Reactor Trip Setpoints & 3.3.1.1, Instrumentation - Reactor Protective Instrumentation
| author name = Bowling M
| author affiliation = NORTHEAST NUCLEAR ENERGY CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000336
| license number = 
| contact person = 
| document report number = B17190, NUDOCS 9807280066
| package number = ML20236T265
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 26
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ - - . - __.                                                           _ _ _ _ _ _        - - - - . - . -  - . _ _ _ _  _ . _ _ _ . . - _
R pe Feny RddRme 156), Watufwd, CT 06385 Northeast Nudrar Energy                                                       Mal, tone Nucica Power Station Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385-0128 (860) 447 1791 Fax (860) 444-4277 The Northeaat Utilities System JUL' 2 l 1998 Docket No. 50-336 B17190 Re: 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protection and Enaineered Safety Features Trio Setooints introduction Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby                                                                    ;
proposes to amend Operating License DPR-65 by incorporating the attached proposed changes into the Technical Specifications of Millstone Unit No. 2. NNECO is proposing to change Technical Specifications 2.1.1, " Safety Limits - Reactor Core," 2.2.1,
                                                                                                          " Limiting Safety System Settings - Reactor Trip Setpoints," 3.3.1.1, " Instrumentation -
Reactor Protective Instrumentation," 3.3.2.1, " Instrumentation - Engineered Safety Feature Actuation System Instrumentation," and to add Technical Specification 3.7.1.8,
                                                                                                          " Plant Systems - Steam Generator Blowdown Isolation Valves." Information will be added to the Bases of the associated Technical Specifications to address the proposed changes.
[   ,
I
                                                                                                                                                                                                                                                   /
                                                                                                    . The proposed changes to Technical Specifications 2.1.2, 3.3.1.1, and 3.3.2.1 are on the same pages (2-4, 3/4 3-4, 3/4 3-16 and 3/4 3-17) which have been proposed to be changed in a separate {{letter dated|date=May 14, 1998|text=letter dated May 14,1998}}.* The proposed change to Technical                                                          ,
Specification Bases Page B 3/4 7-3 is on the same page which has been proposed to be changed in a separate {{letter dated|date=April 6, 1998|text=letter dated April 6,1998}}.* The proposed changes                                                              /
                                                                                                                  ~ 7 L v [,5
* M. L. Bowling letter to the NRC, " Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System Instrumentation," dated May 14,1998.
                                                                                                   '*                 M. L Bowling letter to the NRC, " Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Compliance issues Number 3," dated April 6,1998.
9907290066 990721 PDR                   ADOCK 05000336 P                                    PDR     g

i
       ,       U.S. Nucitar Regulatory Commission                                                              j
   .     -     B17190/Paga 2
                                                                                                              -l I
contained in this letter do not assume approval of any of the previously submitted               l changes.
l Attachment 1 provides a discussion of the proposed changes and the Safety Summary.
Attachment 2 provides the Significant Hazards Consideration. Attachment 3 provides the marked-up version of the appropriate pages of the current Technical Specifications.
Attachment 4 provides the retyped pages of the Technical Specifications.
Environmental Considerations
                                                                                                                )
NNECO has reviewed the proposed License Amendment Request against the criteria                   I of 10CFR51.22 for environmental considerations. The proposed changes to the                     !
I setpoints associated with the Reactor Protection System and the Engineered Safety Features Actuation System will not increase the type and amounts of effluents that may be released off site. In addition, this amendment request will not significantly increase        l individual or cumulative occupational radiation exposures. Therefore, NNECO has determined the proposed changes will not have a significant effect on the quality of the human environment.
Conclusions                                                                                     f The proposed changes w'.sre evaluated utilizing the criteria of 10CFR50.59 and were              i l
determined not to involve an unreviewed safety question. In addition, we have concluded the proposed changes are safe.
l The proposed char ges do not involve a significant impact on public health and safety           i (see the Safety Summary provided in Attachment 1) and do not involve a Significant Hazards Conside;ation pursuant to the provisions of 10CFR50.92 (see the Significant Hazards Consideration provided in Attachment 2).
Plant Operations Review Committee and Nuclear Safety Assessment Board The Plant Operations Review Committee and Nuclear Safety Assessment Board have reviewed and concurred with the determinations.
Schedule We request issuance at your earliest convenience, with the amendment to be                       ,
implemented within 60 days of issuance.
State Notification in accordance with 10CFR50.91(b), a copy of this License Amendment Request is being provided to the State of Connecticut.
l-                                                                                            ._ _ __ _ ___ _

                     . 9                                                                                                                                                                             j
                       ,               U.S. Nucirr Regul tory Commission                                                                                                                                 l
             .           .             B17190/P ge 3 If you should have any questions on the above, please contact Mr. Ravi Joshi at (860) 440-2080.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY M. L. Bowling, Jr.                                              f Recovery Officer - Technical Services Sworn to and subscribed before me thisON dayof TIAa                                                                     1998 knrh 0                          ~~                                        ~~
                                                                                                                      /
otar) PL/bic My Commission expires                                                      / /3 / clDOD Attachments (5) cc:                     H. J. Miller, Region i Administrator D. G. Mcdonald, Jr., NRC Senior Project Manager, Millstone Unit No. 2 D. P. Beaulieu, Senior Resident inspector, Millstone Unit No. 2 W. D. Travers, Ph.D Director - Special Projects Office W. D. Lanning, Deputy Director of Inspections - Special Projects Office P. F. McKee, Deputy Director of Licensing - Special Projects Office Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127 l
l.
i I
L                                            _ _ . _                                                                                         . - - _ _ . - _ _ - _ . . _ _ _ _ _ _ _ _ _ _ _ _ . - _ _

      .-                                              3
                                                                                                                                           ]
  .                                    .                                                                                                   j Docket No. 50-336 B17190 i
I Attachment 1                            j Millstone Nuclear Power Station, Unit No. 2                      i Proposed Revision to Technical Specifications                      l Reactor Protection and Engineered Safety Features Trip Setpoints Discussion of Proposed Changes L
L July 1998

                ,       U. S. Nucler Rrgulatory Commission
    .,            ,     B17190/ Attachment 1/Page 1 Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Features Trip Setpoints Discussion of Proposed Changes introduction
{
Northeast Nuclear Energy Company (NNECO) hereby proposes to amend Operating i                       License DPR-65 by incorporating the attached proposed changes into the Technical                                .
!                       Specifications of Millstone Unit No. 2. NNECO is proposing to change Technical                                  I Specifications 2.1.1, " Safety Limits - Reactor Core," 2.2.1, " Limiting Safety System l
Settings - Reactor Trip Setpoints," 3.3.1.1, " Instrumentation - Reactor Protective i=                      Instrumentation," 3.3.2.1, " Instrumentation - Engineered Safety . Feature Actuation System instrumentation," and to add Technical Specification 3.7.1.8, " Plant Systems -
q l                       Steam Generator Blowdown isolation Valves." Information will be added to the Bases                               1 of the associated Technical Specifications to address the proposed changes.
Description of Proposed Chanoes NNECO has revised the instrument uncertainty and setpoint calculations associated l                       with the Reactor Protection System (RPS) and the Engineered Safety Features f
Actuation System (ESFAS) instrumentation. These calculations include changes to the analytical limits as a result of revisions to the accident analyses, and the effects of a harsh environment (pressure, temperature, and radiation), where appropriate.
The calculation method used has also been changed. The guidelines for establishing                               i i
the analytical limit, trip setpoint, and allowable value are contained in ISA-RP67.04, Part II,1994, " Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation." Several methods for determining the setpoint are discussed                            ;
in ISA-RP67.04. The method used should allow the necessary allowances to account l
for uncertainties between analytical limit, trip setpoint, and allowable value. The previous setpoint calculations utilized " Method #2" of ISA-RP67.04 to determine the trip                         ,
setpoints and allowable values. Recent evaluation has determined that " Method #1" is                            !
more appropriate. This method provides a larger difference between the trip setpoint and the allowable value for the same total instrument loop uncertainty. It will also result in a more conservative trip setpoint. Therefore, the current revision to the setpoint
!.                      calculations use " Method #1."
l                       The calculations also include the effects of a 24 month refueling cycle on instrument
!                     . drift. However, this submittal does not address changing Millstone Unit No. 2 from an 18 month to a 24 month refueling cycle. The use of a larger value for instrument drift is conservative.
The major changes proposed will modify various RPS and ESFAS setpoints and allowable values. Additional minor changes have been proposed to correct various

I                                                                                                      l
   .     \                                                                                             \
     ,     U. S. Nuclxr R:gulatory Commission
: l.     . B17190/ Attachment 1/Page 2                                                                 '
items identified during the review of the Millstone Unit No. 2 Technical Specifications.
Each proposed change is discussed. Additional background information is included, as necessary, to explain the changes. Related changes are grouped together. However, the marked up pages contained in Attachment 3 are sequenced in numerical order by page number. Tables 1 and 2 summarize the proposed changes to the RPS and ESFAS trip setpoints and allowable values.
!          RPS Setpoint Changes Technical Specification 2.2.1 Table 2.2-1
: 1. The trip setpoint in Table 2.2-1, " Reactor Protective Instrumentation Trip   I Setpoint Limits," for Pressurizer Pressure - High, Functional Unit 5, will be changed from 5 2400 psia to 5 2397 psia. The allowable value will be changed from 5 2408 psia to 5 2407 psia. These changes are the result
,                        of revisions to the instrument uncertainty and setpoint calculations. The analytical limit did not change and the effects of a harsh environment have not been included.
I The RPS actuation on high pressurizer pressure, in addition to causing a reactor trip, will also open both pressurizer power operated relief valves f
(PORVs) to help mitigate the high pressure transient.
: 2. The trip setpoint in Table 2.2-1 for Containment Pressure - High,             1 Functional Unit 6, will be changed from 5 4.75 psig to 5 4.42 psig. The       f allowable value will be changed from 5 5.24 psig to 5 5.07 psig. These changes are the result of revisions to the instrument uncertainty and setpoint calculations. The analytical limit did not change and the effects of a harsh environment have not been included.
: 3. The trip setpoint in Table 2.2-1 for Steam Generator Pressure - Low, Functional Unit 7, will be changed from 2 680 psia to 2 691 psia. The allowable value will be changed from 2 672 psia to 2 677 psia. These changes are the result of revisions to the instrument uncertainty and setpoint calculations. The analytical limit for the closure of a main steam line isolation valve (MSIV) did not change, and the effects of a harsh environment have not been included.
This RPS actuation is also used to mitigate a main steam line pipe break.
It is the backup to the high containment pressure RPS actuation for a main steam line break inside containment. The total loop uncertainty for      j i                        the RPS actuation on low steam generator pressure following a main            I l                        steam line break includes the effects of a harsh environment (high temperature and pressure). This increases the total loop uncertainty to approximately 94 psi. This results in the need to use an analytical limit of l

     .                           s
       ,                           U. S. Nuctur Regulatory Commission
  .-                           . B17190/ Attachment 1/Page 3 i
                                                < 597 psia (691 psia - 94 psia) to ensure that the RPS setpoint for MSIV closure is bounding. The analytical limit used in the analysis of this event has been conservatively increased from 478 psia to 550 psia. However, this analytical limit and resultant setpoint, t 644 psia, are bounded by the proposed RPS trip setpoint of 3 691 psia for the MSIV closure event.
Therefore, the RPS actuation on low steam generator pressure will occur above the value assumed in the main steam break analysis.
: 4.   - The trip setpoint in Table 2.2-1 for Steam Generator Water Level - Low, Functional Unit 8, will be changed from 2 36.0 % to 139.1 %. The allowable value will be changed from 135.2 % to 138.0 %. These changes are the result of revisions to the instrument uncertainty and setpoint calculations. The analytical limit did not change and the effects i
of a harsh environment have not been included.
: 5. The value of steam generator pressure contained in Table 2.2-1, Table Notation (2) will be changed from 780 psia to 800 psia. This change is to      ,
restore the margin to bypass the automatic reactor trip on low steam generator pressure. N proposed increase in the actuation setpoint from "2 680 psia" to "E 691 psia" reduces the margin for manual bypassing of the reactor trip.
In addition, the words "when steam generator pressure is < 800 psia and" will replace "below 780 psia when," and "when steam generator pressure is t 800," will replace "at or above 780 psia." The addition of the plant specific parameter is not a technical change. Table Notation (2) only applies to the reactor trip on low steam generator pressure. Therefore, l                                               the plant parameter to use to determine when this RPS actuation can be bypassed is steam generator pressure. This change will only improve the clarity of the statement. These changes are consistent with the format of Table 3.3-3 Table Notation (a) and will not result in any technical change, except as previously discussed.
l Technical Specification 3.3.1.1 Table 3.3-1 L
: 1. The value of steam generator pressure contained in Table 3.3-1, " Reactor Protective Instrumentation," Table Notation (b) will be changed from 780 psia to 800 psia. This change is to restore the margin to bypass the automatic reactor trip on low steam generator pressure. The proposed increase in the actuation setpoint from "2 680 psia" to "E 691 psia" reduces the margin for manual bypassing of the reactor trip.
in addition, the words "when steam generator pressure is < 800 psia and" will replace "below 780 psia when," and "when steam generator pressure         )
is t 800," will replace "at or above 780 psia." The addition of the plant       l

s'          )
      ,                       U. S. Nucitar Regulatory Commission
   .    .                    . B17190/ Attachment 1/Pege 4 spec;fic parameter is not a technical change. Table Notation (b) only applies to the reactor trip on low steam generator pressure. Therefore, the plant parameter to use to determine when this RPS actuation can be bypassed is steam generator pressure. This change will only improve the clarity of the statement. These changes are consistent with the format of Table 3.3-3 Table Notation (a) and will not result in any technical change, except as previously discussed.
ESFAS Setpoint Changes Technical Specification 3.3.2.1 Table 3.3-4 l                                    1. The trip setpoint in Table 3.3-4, " Engineered Safety Feature Actuation System Instrumentation Trip Values," for Containment Pressure - High, Functional Units 1.b., 3.c., 4.b., and 5.c. will be changed from 5 4.75 psig to 5 4.42 psig. The allowable value will be changed from 5 5.20 psig to 5 5.07 psig. These changes are the result of ravisions to the instrument uncertainty and setpoint calculations. Tiie analytical limit did not change and the effects of a harsh environmen: have not been included.
1 In addition a "5" will be added to the trip setpoint_ for Functional Unit 1.b.
l                                           The addition of this symbol is consistent with Functional Units 3.c., 4.b.,
and 5.c. This will correct an error on Technical Specification Page 3/4 3-18. The "5" symbol was inadvertently omitted by NNECO in the license amendment request dated October 28,1992.N This license amendment request included an incorrect Technical Specification Page 3/4 3-18. This led to the error on Page 3/4 3-18 when License Amendment No.167* to Facility Operating License DPR-65 was issued. The elimination of the "5" symbol was an administrative error and was not discussed or justified in the Safety Evaluation Report for License Amendment No.167.
: 2. The trip setpoint in Table 3.3-4 for Pressurizer Pressure - Low, Functional Units 1.c., 3.d., and 5.d. will be changed from 21600 psia to 21714 psia.
The allowable value will be changed from 21592.5 psia to 21704 psia.
These changes are the result of revisions to the instrument uncertainty and setpoint calculations. The definition of the analytical limit for the application of the setpoint methodology has been increased from 1578 psia to 1600 psia and the effects of a harsh environment (pressure,
* J. F. Opeka letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2, Proposed Revision to Technical Specifications, Main Steam Line j:                                   Break Design Limits," dated October 28,1992.
* G. S. Vissing letter to J. F. Opeka, " Issuance of Amendment (TAC NO. M84774)," dated December 23,1992.

    . s
      ,     U. S. Nucinar Reguintory Commission
  .     . B17190/Attachm:nt 1/Prgs 5 temperature, and radiation) have been included.                                            This is a more conservative change to the analytical limit (i.e., the analytical limit will be reached sooner during the event). It should be noted that a nominal value is used in the loss of coolant accident analysis.
: 3. The trip setpoint in Table 3.3-4 for Steam Generator Pressure - Low, Functional Unit 4.c. will be changed from 2 500 psia to 25 72 psia. The allowable value will be changed from 1492.5 psia to 1558 psia. These changes are the result of revisions to the instrument uncertainty and setpoint calculations. The analytical limit has not changed. The effects of a harsh environment (pressure and temperature) have been included.
: 4. The trip setpoint in Table 3.3-4 for Refueling Water Storage Tank Level -
Low, Functional Unit 6.b. will be changed from 48 9 inches to 4613 inches. The allowable value will be changed from 48118 inches to 461 6 inches. These changes are the result of revisions to the instrument uncertainty and setpoint calculations. The maximum analytical limit (66 inches) has not changed. A lower analytical limit (26 inches) has been added. The change in the setpoint from 48 inches to 46 inches was done to select an RWST level at the midpoint between the minimum and maximum analytical limits. The effects of a harsh environment have not been included.
: 5. The trip'setpoint in Table 3.3-4 for Steam Generator Level - Low, Functional Unit 9.b. will be changed from 212 % to 2 26.8 %. The
                                   ~
allowable value will be changed from 210 % to 2 25.2 %. These changes are the result of revisions to the instrument uncertainty and setpoint calculations. The analytical limit for harsh environment events has been increased from 0 % to 5 %. The effects of a harsh environment (pressure, temperature, and radiation) have been included.                                            This is a more        ;
conservative change to the analytical limit (i.e., the analytical limit will be                                  1 reached sooner during the event). It also bounds the analytical limit for the non-harsh environment events.
Technical Specification 3.3.2.1 Table 3.3-3
: 1. The value of pressurizer pressure contained in Table 3.3-3 Table Notation (a) will be changed from 1750 psia to 1850 psia. This change is I                        necessary to allow sufficient margin to block the automatic ESF actuation on low pressurizer pressure. The proposed increase in the actuation setpoint from 21600 psia to 21714 psia will not provide sufficient margin for the manual blocking of the ESF actuations unless the pressure when blocking is permitted is increased.
1
__ __-_____- -_________                __ __a

    ,                                   i
{
      ,,                                       U. S. Nuciter Regulatory Commission
  ,                       ,                    B17190/Attrchmtnt 1/Pcg3 6 The value of pressurizer pressure used to define Mode 3* for Technical
;                                                            Specifications 3.5.1, " Emergency Core Cooling Systems (ECCS) - Safety                                     1 L                                                            Injection Tanks," 3.5.2, " Emergency Core Cooling Systems - ECCS l                                                            Subsystems - Tavg 2 300 *F," 3.5.3, " Emergency Core Cooling Systems -                                     1 ECCS Subsystems - Tavg < 300 *F," and 3.6.2.1, " Containment Systems                                       I
                                                             - Depressurization and Cooling Systems Containment Spray and Cooling Systems," will not be changed. It will remain at 1750 psia. The
;                                                            operability of the safety injection tanks, the emergency core cooling l                                                           . subsystems, and the containment spray system is not affected by blocking the low pressurizer. pressure ESF actuations (safety injection, l                                                            containment isolation, and enclosure building filtration). In Mode 3, even
!                                                            when the low pressurizer. pressure ESF actuation can be blocked, the j                                                            high containment pressure ESF actuation is required to be operable.
l Also, in Modes 3 and 4 manual ESF actuation capability is required to be operable. Therefore, blocking the low pressurizer pressure ESF actuations does not affect the operability of the safety injection tanks, the emergency core cooling subsystems, and the containment spray system.
i
: 2. The value of steam generator pressure contained in Table 3.3-3 Table                                       !
Notation (c) will be changed from 600 psia to 700 psia. This change is                                     {
necessary to allow sufficient margin to block the automatic ESF actuation                                  l l                                                            on low steam generator pressure. The proposed increase in the actuation setpoint from 2 500 psia to 1572 psia will not provide sufficient margin for                               I the manual blocking of the ESF actuation unless the pressure when blocking is permitted is increased.
in addition the words " steam generator pressure" will be added to indicate what plant parameter is used, "<" will replace "below," and "2" will replace "at or above." 3e addition of the plant specific parameter is not a tech ical chanW lable Notation (c) only applies to the main steam line isuation signal a low steam generator pressure. Therefore, the plant parameter to use to determine when this ESF function can be bypassed is steam generator pressure. This change will only improve the clarity of the statement. These changes are consistent with the format of Table 3.3-3 L                                                            Table Notation (a).
: 3. The value of pressurizer pressure contained in Table 3.3-3 Action 2 parts L                                                            a and b and Action 4 parts a and b will be changed from 1750 psia to 1850 psia. The distinction between action statement requirements
;                                                            appears to be based on whether the ESF actuation on low pressurizer L                                                            pressure is bypassed or not. Such a distinction is not technically required, and is not included in either NUREG - 0212, " Standard Technical Specifications for Combustion Engineering Pressurized Water f                                                            Reactors," Revision 2, or in NUREG - 1432, " Standard Technical Specifications Combustion Engineering Plants," Revision 1.                               However,

                         ,                                                      U. S. Nucitar Regulatory Commission
          .                                      .                              B17190/ Attachment 1/Pcg3 7 this change will keep the numbers in this table consistent, which will eliminate- any possible confusion by the operating crew when implementing the action statement requirements. This proposed change is consistent with the change to Table 3.3-3 Table Notation (a) previously discussed.
Steam Generator Blowdown Isolation Changes Technical Specification 3.7.1.8 Technical Specification 3.7.1.8, " Steam Generator Blowdown Isolation Valves,"      will be added. Automatic isolation of steam generator blowdown is assumed in the new loss of main feedwater analysis.
Therefore, the operability of the steam generator blowdown isolation valves should be required by Technical Specifications. Isolation of steam generator blowdown will occur at the same steam generator level as         l auxiliary feedwater (AFW) actuation.                                       '
The proposed action and surveillance requirements are consistent with Technical Specification 3.6.3.1, " Containment isolation Valves." The steam generator blowdown isolation valves are also containment isolation valves.
Technical Specification Index Technical Specification Index Page Vill will be changed to reflect the addition of Technical Specification 3.7.1.8, " Steam Generator Blowdown Isolation Valves."
Technical Specification 3.3.2.1 Table 3.3-3 Functional Unit 10, " STEAM GENERATOR BLOWDOWN," will be added to Table     3.3-3, " Engineered Safety Feature Actuation System Instrumentation." Automatic isolation of steam generator blowdown is assumed in the new loss of main feedwater analysis. Isolation of steam generator blowdown will occur at the same steam generator level as AFW actuation (Functional Unit 9.b). The requirements for this actuation are the same as the automatic AFW actuation.
l l                                                                                                                                                                                1 l
l
                                                    -                           __      -_..___m.____

    .       s
      ,       U. S. Nucl: r R:gulctory Commission B17190/Attachm:nt 1/Pcge 8 Technical Specification 3.3.2.1 Table 3.3-4 Functional Unit 10, " STEAM GENERATOR BLOWDOWN," will be added to Table 3.3-4. Automatic isolation of steam generator blowdown is assumed in the new loss of main feedwater analysis. Isolation of steam generator blowdown will occur at the same steam generator level as AFW actuation (Functional Unit 9.b). The requirements for this actuation are the same as the automatic AFW actuation.
Technical Specification 3.3.2.1 Table 4.3-2 3
Functional Unit 10, " STEAM GENERATOR BLOWDOWN," will be added to Table ' 4.3-2, " Engineered Safety Feature Actuation System                          i instrumentation Surveillance Requirements." Automatic isolation of steam                j generator blowdown is assumed in the new loss of main feedwater analysis. Isolation of steam generator blowdown will occur at the same steam generator level as AFW actuation (Functional Unit 9.b). The                       J requirements for this actuation are the same as for automatic AFW actuation.
Technical Specifications 3/4.3.1 and 3/4.3.2 Bases A discussion of the automatic isolation of steam generator blowdown on                  :
low steam generator level will be added.
J Technical Specification 3.7.1.8 Bases L                                                    A discussion of the steam generator blowdown isolation valves will be                   j added.                                                                                  j The proposed change to add the automatic isolation of steam generator blowdown on low steam generator level will require modification of the auxiliary feedwater actuation and steam generator blowdown isolation circuits. The plant modification has been evaluated in accordance with 10 CFR 50.59. The L                   evaluation determined that the modification is safe and does not involve an unreviewed safety question. The components added will be seismically supported and qualified Class 1E. They have the same reliability as the existing seismically supported, Class 1E components in the auxiliary feedwater control system and blowdown isolation circuits. The added logic is redundant so no single failure can cause the logic to fail to function. The added logic is in the form of an OR logic with the existing system logic so no operation or failure of the added feature can prevent the existing logic from performing its intended function.
i t

    .       s
      ,              U. S. Nuclear Regulatory Commission
  .       .          B17190/ Attachment 1/Page 9 Miscellaneous Changes Technical Specification 2.1.1 Figure 2.1-1 l                                    The value of thermal power stated on Figure 2.1-1, " Reactor Core Thermal Margin Safety Limit - Four Reactor Coolant Pumps Operating,"
that is limited by the high power level trip will be changed from "112%" to i
                                     "111.6%." This is the value used in the safety analysis. This is consistent with the maximum allowed high power trip setpoint of 106.6%, plus 5%
unce.dnty.
A high power trip value of 107%, plus 5% uncertainty, is consistent with the original desgn basis of Millstone Unit No. 2. However, NNECO stated in the submittal to support Cycle 4 operation that the RPS setpoints had been updated. This submittal proposed to change the maximum value of the high power trip from 107% to 106.6%. This prog)osed change was approved and issued as License Amendment No. 61.              The high power trip setpoint has not been changed since License Amendment No. 61 was approved.
The value of 106.6% should appear with the trip setpoint for Functional Unit 2, " Power Level High," in Technical Specification Table 2.2-1,
                                    " Reactor Protective Instrumentation Trip Setpoint Limits." The value of 106.6% was inadvertently deleted and a request to add it back to this table has been submitted by NNECO.8)
Technical Specification 2.2.1 Table 2.2-1
: 1. The word "lS" in Table 2.2-1 Notation (3) will be changed to "is." This is not a defined word. It should not be capitalized.
: 2. The word "steram" in Table 2.2-1 Notation (5) will be changed to " steam."
The current word is not spelled correctly.
W. G. Counsil letter to the NRC,
* Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications," dated August 29,1980.
* R. A. Clark letter to W. G. Counsil, License Amendment No. 61 to Facility Operating License DPR-65, dated October 6,1980.                                                   l
* M. L. Bowling, Jr. letter to the NRC, " Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Feature Actuation System Instrumentation," dated May 14,1998.

s             ' *f
       ,                      . U. S. Nucl:ar Regulatory Commission B17190/ Attachment 1/Page 10 Technical Specification 2.1.1 - Bases The maximum reactor power level will be changed from 112% to 111.6%.
This is the value currently used in the safety analysis. This change has already been discussed.
Technical Specification 2.1.2 - Bases The word "PESSURE" in the title will be changed to " PRESSURE." The current word is not spelled correctly.
Technical Specification 2.2.1 - Bases
: 1. Power Level- High The maximum reactor power level will be changed from 112% to 111.6%.
This is the value currently used in the safety analysis. This change has already been discussed.
: 2. Reactor Coolant Flow - Low The reference to power operation with less than both Reactor Coolant System loops and all four reactor coolant pumps in operation will be removed. Power operation is not allowed with less than two steam generators and four reactor coolant pumps in operation.
: 3.      Pressurizer Pressure - High The word "approximately" will be added since the proposed trip setpoint is not exactly 100 psi from the nominal lift setpoint of the pressurizer safety valves.
: 4.      Steam Generator Pressure - Low The phrase "The setting of 680 psia" will be changed to "The trip setting" to be consistent with the proposed trip setpoint and since it is not l                                            necessary to repeat the Technical Specification setpoint in the Bases.
l                                            The last sentence, which discusses the uncertainty factor, will be deleted.
This information is already contained in the FSAR. It is not necessary for the information to be repeated in the Bases.

      ,-                                                  U.'S. Nucimr Regulitory Commission
  ,      ..                                               B17190/ Attachment 1/Page 11                                                                                                                                        4
: 5. Thermal Margin / Low Pressure The value of temperature measurement uncertainty will be changed from "2 *F" to "2.25 'F." This is the correct value for this instrumentation and is consistent with the value specified in the FSAR.
6.'     Loss of Turbine The word " values" will be changed to " valves" in the Loss of Turbine discussion. The current word is not spelled correctly.
Technical Specification 3.3.2.1
: 1. The word "VALUE" will be replaced with "SETPOINT" for the heading of the second column of Table 3.3-4 on Page 3/4 3-19. This is consistent I                                                                          with the column heading on the other pages associated with this table.
: 2. The amendment numbers for Table 3.3-4 Page 3-19 will be added. This l                                                                          page was previously changed by License Amendments 45,49, and 72.
!                                                       . Safety Summarv
!,                                                      The proposed changes include:
: 1.        Corrections to the specified maximum reactor power level limited by the high i                                                                   power level RPS trip.
;                                                       2.          RPS trip setpoint, allowable value, and bypass setpoint changes.
l 3          ESFAS trip setpoint, allowable value, and block setpoint changes.
: 4.          Addition of a new Technical Specification and additional requirements associated with the automatic isolation of steam generator blowdown.
: 5.         Minor editorial and non-technical changes to correct spelling errors, correct a
                                                                  . capitalization error, add page amendment numbers, add the specific plant parameter (steam generator pressure) to use if an ESF function can be bypassed, change the value of the parameter (pressurizer pressure) used in g                                                                   action statements, add a ",s" symbol, change "value" to "setpoint," and update the index.-
: 6.          Revisions to the Bases of the Technical Specifications to incorporate the RPS
                                                                  ' and ESFAS setpoint changes, correct errors identified during the review of the Millstone Unit No. 2 Technical Specifications, eliminate redundant information, i
1
       -    _ _ - - - - - - _ - - - - - - - _ - _ - _ - - - - - -                               --   ---- - - - - - - - _ - - - - - - - - - - -                                                           -------------------d

U. S. Nucl ar Regulatory Commission
  .                        . B17190/Att chment 1/Pcge 12 and expand the Bases to discuss the new requirements for steam generator blowdown isolation.
The proposed change to correct the maximum reactor power level from 112% to 111.6% is consistent with the maximum high power trip setpoint of 106.6%, plus 5%
uncertainty, currently used in the safety analyses. This does not change the current Technical Specification maximum high power reactor trip setpoint. Therefore, the RPS will continue to function as before.
The proposed changes to the trip setpoints and allowable values for the RPS trips on high pressurizer pressure, high containment pressure, low steam generator pressure, and low steam generator level are the result of revisions to the instrument loop uncertainty and setpoint calculations. These calculations were revised to incorporate calculation methodology changes, analytical limit changes, correct errors identified, and to include the effects of a harsh environment (pressure, temperature, and radiation), where appropriate. The proposed setpoints and allowable values will ensure a reactor trip signal is generated at, or before the analytical limits used in the respective accident analyses are reached. In addition, the proposed changes will not result in any significant restriction to plant operation. The setpoint changes are small, and will r'ot significantly reduce the margin between normal plant operation and RPS actuation. Therefore, the RPS will continue to function as before.
The proposed changes to the trip setpoints and allowable values for the ESFAS actuations on low pressurizer pressure, high containment pressure, low steam generator pressure, low refueling water storage tank level, and low steam generator level are the result of revisions to the instrument loop uncertainty and setpoint calculations. These calculations were revised to incorporate calculation methodology changes, analytical limit changes, correct errors identified, and to include the effects of a harsh environment (pressure, temperature, and radiation), where appropriate. The proposed setpoints and allowable . values will ensure an ESF actuation signal is generated at, or before the analytical limits used in the respective accident analyses are reached. In addition, the proposed changes will not result in any significant restriction to plant operation. The setpoint changes are small, and will not significantly reduce the margin between normal plant operation and ESFAS actuation. Therefore, the ESFAS will continue to function as before.
The proposed changes to add Technical Specification requirements for the steam generator blowdown isolation valves 'will provide additional assurance that the automatic isolation of steam generator blowdown will occur as assumed in the loss of l
                             ' main feedwater accident analysis. Therefore, the ESFAS will function as assumed in l                              the accident analysis.
l f                              The proposed change to the value of steam generator pressure when the steam I                              generator low pressure reactor trip can be bypassed (from 780 psia to 800 psia ) will reduce the range of plant operation when this trip is required to be available. However,
    -_ _ _ _ _ _ - -                                        __ _         . _ _ _ _ _ _ .   . _ _ _ _ - - - . _ _ _ - - _ _ _ _ _ _ _ _____ A

         ,                           U. S. Nucle:r Regulatory Commission B17190/ Attachment 1/Page 13 this will not affect the range of plant operation when this RPS trip is required to be operable. This RPS trip is required in Modes 1 and 2. The expected steam generator pressure during a reactor startup (entry into Mode 2) is approximately 900 psia, which corresponds to a Raactor Coolant System (RCS) temperature of approximately 532 *F.
The proposed change will require the bypass to be automatically removed prior to exceeding a steam generator pressure of 800 psia. Therefore, the RPS will continue to function as before.
The proposed change to the value of pressurizer pressure (from 1750 psia to 1850 psia) when the pressurizer low pressure ESF actuations (SIAS, CIAS, and EBFAS) can be blocked will reduce the range of plant operation when these functions are required to be available. However, since the plant would normally be in Mode 3 when pressurizer pressure is in this range, automatic actuation of these ESF functions on high containment pressure, as well as manual actuation, is required to be operable. In addition, the plant would not normally maintain pressurizer pressure betwetni 1750 psia and 1850 psia. _ Below 1750 psia only one train of emergency core cooling subsystems is required to be operable, and no containment spray trains are required to be operable. During a plant shutdown the pressurizer pressure will normally be reduced to below 1750 psia to relax the Technical Specification requirements for emergency core cooling subsystems and containment spray. During a plant startup pressurizer pressure will normally be maintained below 1750 psia until the necessary emergency core cooling and containment spray equipment has been declared operable. Once pressurizer pressure exceeds 1750 psia during a plant startup, the increase normally continues to the normal operating pressure of approximately 2250 psia. Therefore, since automatic actuation of these ESF functions on high containment pressure, as well as manual actuation,'should be operable, and the time the plant will operate between 1750 psia and 1850 psia is small, the ESFAS will continue to function as before.
I The proposed change to the value of steam generator pressure (from 600 psia to 700 psia) when the steam generator low prossure ESF actuation (main steam iine isolation)
;                                    can be blocked will reduce the range of plant operation when this function is required to l                                    be available. However, since the plant would be in Mode 3 when steam generator pressure is in this range (RCS temperature of approximately 486 F to 503 F),
automatic actuation of this ESF function on high containment pressure, as well as manual actuation, is required to be operable. In addition, the plant would not normally maintain steam generator pressure between 600 psia and 700 psia. Therefore, since i                                    automatic actuation of this ESF function on high containment pressure, as well as manual actuation, should be operable, and the time the plant will operate between 600 psia and 700 psia is small, the ESFAS will continue to function as before.
The minor editorial and non-technical changes to correct spelling errors, correct a               ;
capitalization error, add page amendment numbers, add the specific plant parameter                ]
(steam generator pressure) to use if an RPS or ESF function can be bypassed, change the value of the parameter (pressurizer pressure) used in action statements, add a '5" symbol, change "value" to "setpoint," and update the index will have no effect on plant           i I

1 I
      .                               s j            ,                          U. S. Nucl r Regul: tory Commission 3
817190/Attechm:nt 1/Page 14 operation. These changes will not result in any technical changes to the Millstone Unit l
No. 2 Technical Specifications. Therefore, the RPS and ESFAS will continue to 4'
l                                       function as before.
l                                       The proposed changes to the Technical Specification Bases will incorporate the RPS and ESFAS setpoint changes, correct errors, eliminate redundant information, and expand the Bases to discuss the new requirements for steam generator blowdown isolation. These changes will have no effect on equipment operation. Therefore, the RPS and ESFAS will continue to function as before.
The proposed changes have no adverse effect on any design basis accident previously evaluated and have no adverse effect on how the RPS and ESFAS function to mitigate the consequences of design basis accidents. Therefore, there is no adverse impact on public health and safety.
1 i

rl         2      5       6          6 PA           5       $       1          1                   2 e     a     ig        a          a t l nbe is p       s p
i s
p i
s                 %
eau r          8 p
2 r wla      0 4       2          2                   5 2       7 uoV Cl   l 4
2      5       6 7
6                   3 A               5       1          1                   1 5
dt         i a       g       a          a en          s     is i
s i
s                 %
si          p       p       p          p o
p op      7 9
2       1          4 1
t             4       9                              9 oe        2 3
4       6 4
6                   3 PrS          5       5       1          2                   2 t     i a       g       a          a t            s ni n i          i is        s          s s               p       p       p          p                 %
e n
eo rp r
0       5       0          0                   6 n             0       7       8          8                   3 ut e       4 a                     4       6          6                     1 h    CS         2 5       2          2 C                 5 t
n                                              d i
o   y                                           n p  t           t n
t n
t n
t a            t t
e i ns          e       e       e n
e     .
p s.
n e
S    ta r e c
t m       m u
m          m    ms             m S     ef f u
r       r u
r u
r    ee r u
r P     cE n
t s
t s
t s
t s   T P           t s
R                 n       n     in           n   h               n 1
U           i       I                  I ig            I H
E L
B A   dl a                                     a   dl     7 T     e ct si      i e       e       e        i s
e  a r ic 9
e otym        m       m       m          p   iut y 5 )a m pl o
i   a       a       a        0      ql e a ipS < is a r naL t
S       S       S          5 n          PA                                     5     Rnm
(   ai i
o                                                              l s
is                                               )
m                                           et r n m5 o    1      t a l
i a     ig        a  uev     an e
nct         s             is Ceg            ei          p       s       s  oE    isLik  a            %
r ym t    i p       pl         p ya r            rl        2       3       8 Cg n 8   mer             4 oP/
i uaL       2       8       5Vi        7   aB              3 Cn          4                          4t e t
la1               A      2       5       6ISm  it S
u gen t
mil e                                              (
Rmr c h                      h       h           w aa                       r gngt
_   et   t ei zHeH i
r o oL r
o w l
cA           p           r iu e m       mtae                           mt araL o
u/   0     i r in  r e ar r N. 19        T r
suau s st        s eens t       u                      t eel n ev es   nsSes                                   Se e S7                         r eoe             Ger                           GL UB
      .1 PrCr P       P          P
-                                                                                  l l:

1 PA            2                 5           2                              2 e                      ig          ia t l         5                   s           s nb e                           p           p            8    s          %
eau r
2a 9i                           5 1
e r wla      5 sp              0                            1hc           0 2           2 uoV Cl   l 1
5           9              8 n 4i 1
2 A        2                           4 5           2 dt                            ig            a en si          4                  s p
i s
p 0 s             %
o p op 1 a 7i s              2           2 3 e h
8 oe  t       1   p            4           7                1c            6 2
2               4           5              6n PrS                              5           1 4 i               2 t    t                        ig            a s   nin        0                   s p         isp            9 s             %
e   eo r p r
0a 6i s              5 e
h          2 n   ut e       1    p            7 0
8 cn n                2 0
5              4i 1
h a CS                             4 2
2 5
C t
i n   y                    .                             .
o               t                                                                     .
t t
n      p s. n             t p s. t                t ps   .
D   ns           e      ms     o n
e n
e      m s n
e n
e      ms e er                          e e i
ta c                                                                              e er t         t m           it e                              a   m           m     T P urm r                    m S     r e ef f u
r T P    i d
u r
u r                              u r
T P S     cE t
s    hg d     a  t s
t s
h   dt                 t hg d A     n            n    in     R     n         In      ig n sn                  s n     in S   U            I Ha         I H aI                   I Ha E
2 E
L   dl a           a B     e ct       is                  e           e mse      m s
e A     si           p                                            uh       u h mc i
T     otym pl      i 0 m
a m
a            in       i m c%
n xi         i5 oaL r n 0
6                 S           S                a      in PA          1                                             M66      M   6 2
n i
o s
i s
m 6             l e
m1 o
t nct a
is a               ig s
ani mse p                           isL     ka       uh Ceg            ei r ym                                                      mcn i
t                           p           p rl        8                                    mer                      %
ya                   i                    3           8              ixi             0 r            uaL       7                  8                 aB         a 6 to/P        Cn A
5 1                  5 7
4t e           M6 la1                                                         S un ge t
e Rm     h                    r  w            t   hS I
w     l r c                       eoS               n gM            r o        e                  r aa                        zLAS              eH i
oL         v e                  o w et   t m e S S mta                     S          mt o ar aLW i       I p             r e lcA u/         i r
urCA s u       F     in    rA Aar eI F          rS L wA auCBteeuM                Stor              eeleF I
0                     ssS B N9           T esAE t     s           ns            LS         Sen evA t
S7
    . 1 prsr eI              nsS E Ses oe            Ger        W                   GL UB
    .1 P             CrA PIS             P      R ll

         .                         s o
Docket No. 50-336 B17190 Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Features Trip Setpoints Significant Hazards Consideration i
l July 1998 r
i

       .        ~
         ;             U. S. Nuclur Rsgulatory Commission
     .       .i        B17190/ Attachment 2/Pcg31-Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Features Trip Setpoints Significant Hazards Consideration Significant Ha7ards Consideration in accordance with 10CFR50.92, NNECO has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve an SHC because the changes
                     .would not:
: 1.                 Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to correct the maximum reactor power level from 112% to 111.6% is consistent with the maximum high power trip sotpoint of 106.6%, plus 5% uncertainty, currently used in the safety analyses. This does not change the Technical Specification required high power reactor trip setpoint. There will be no adverse effect on any design basis accident previously evaluated or on any equipment important to safety. Therefore, the proposed change will not result in a' significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes to the trip setpoints and allowable values for the Reactor Protection System (RPS) trips on high pressurizer pressure, high containment pressure, low steam generator pressure, and low steam generator level are the result of revisions to the instrument loop uncertainty and setpoint calculations. 1 These calculations were revised to incorporate calculation methodology              l changes, analytical limit changes, correct errors identified, and to include the effects of a harsh environment (pressure, temperature, and radiation), where appropriate. The proposed setpoints and allowable values will ensure a reactor trip signal is generated at, or before the analytical limits used in the respective l                                        accident analyses are reached. There will be no adverse effect on any design L                                       basis accident previously evaluated or on any equipment important to safety.
L Therefore, the proposed changes will not result in a significant increase in the probability or consequences of an accident previously evaluated.
l The proposed changes to the trip setpoints and allowable values for the Engineered Safety Features Actuation System (ESFAS) actuations on low
                                 . pressurizer pressure, high containment pressure, low steam generator pressure, low refueling water storage tank level, and low steam generator level are the result of revisions to the instrument loop uncertainty and setpoint calculations.
These calculations were revised to incorporate calculation methodology changes, analytical limit changes, correct errors identified, and to include the L                    1 j

                    ~
l
     .,               U. S. Nucl=r R:gulttory Commission B17190/Attrchment 2/Page 2 i
effects of a harsh environment (pressure, temperature, and radiation), where         l appropriate. The proposed setpoints and allowable values will ensure an ESF actuation signal is generated at, or before the analytical limits used in the respective accident analyses are reached. There will be no adverse effect on any design basis accident previously evaluated or on any equipment important to safety. Therefore, the proposed changes will not result in a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes to add Technical Specification requirements for the steam generator blowdown isolation valves will provide additional assurance that the automatic isolation of steam generator blowdown will occur as assumed in the loss of main feedwater accident analysis. There will be no adverse effect on any design basis accident previously evaluated or on any equipment important to safety. Therefore, the proposed changes will not result in a significant increase in the probability or consequences of an accident previously evaluated.                                                                           j 1
The proposed change to the value of steam generator pressure when the steam generator low pressure reactor trip can be bypassed (from 780 psia to 800 psia) will reduce the range of plant operation when this trip is required to be available. ;
However, this will not affect the range of plant operation when this RPS trip is     {
required to be operable. This RPS trip is required in Modes 1 and 2. The expected steam generator pressure during a reactor startup (entry into Mode 2) is approximately 900 psia, which corresponds to a Reactor Coolant System (RCS) temperature of approximately 532 *F. The proposed change will require the bypass to be automatically removed prior to exceeding a steam generator pressure of 800 psia. There will be no adverse effect on any design basis accident previously evaluated or on any equipment important to safety, lherefore, the proposed change will not result in a significant increase in the probability or consequences of an accident previously evaluated.
!                           The proposed change to the value of pressurizer pressure (from 1750 psia to l                           1850 psia) when the pressurizer low pressure ESF actuations (SIAS, CIAS, and L                           EBFAS) can be blocked will reduce the range of plant operation when these
!                           functions are required to be available. However, since the plant would normally be in Mode 3 when pressurizer pressure is in this range, automatic actuation of these ESF functions on high containment pressure, as well as manual actuation,
!                           is required to be operable. In addition, the plant would not normally maintain pressurizer pressure between 1750 psia and 1850 psia. Therefore, since automatic actuation of these ESF functions on high containment pressure, as well as manual actuation, should be operable, and the time the plant will operate between 1750 psia and 1850 psia is small, the ESFAS will continue to function as before. There will be no adverse effect on any design basis accident previously evaluated or on any equipment important to safety. Therefore, the L       _ _ _ - - _

i l
j
     ;      U. S. Nucinr Regulatory Commission                                                                               I j.
B17190/ Attachment 2/Page 3 t
proposed change will not result in a significant increase in the probability or consequences of an accident previously evaluated.
(
The proposed change to the value of steam generator pressure (from 600 psia
,                  to 700 psia) when the steam generator low pressure ESF actuation (main steam
!                  line isolation) can be blocked will reduce the range of plant operation when this j                 function is required to be available. However, since the plant would be in Mode
!                  3 when steam generator pressure is in this range (RCS temperature of j                  approximately 486 *F to 503 F), automatic actuation of this ESF function on high containment pressure, as well as manual actuation, is required to be operable. In addition, the plant would not normally maintain steam generator i
pressure between 600 psia and 700 psia. Therefore, since automatic actuation of this ESF function on high containment pressure, as well as manual actuation, should be operable, and the time the plant will operate between 600 psia and 700 psia is small, the ESFAS will continue to function as before. There will be no adverse effect on any design basis accident previously evaluated or on any                            t equipment important to safety. Therefore, the proposed change will not result in
;                  a significant increase in the probability or consequences of an accident previously evaluated.
The minor editorial and non-technical changes to correct spelling errors, correct a capitalization error, add page amendment numbers, add the specific plant parameter (steam generator pressure) to use if an RPS or ESF function can be I
bypassed, change the value of the parameter (pressurizer pressure) used in action statements, add a "5" symbol, change "value" to "setpoint," and update the index will have no effect on plant operation. These changes will not result in any technical changes to the Millstone Unit No. 2 Technical Specifications.
There will be no adverse effect on any design basis accident previously                                    )
evaluated or on any equipment important to safety. Therefore, the proposed change will not result in a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes to the Technical Specification Bases will incorporate the RPS and ESFAS setpoint changes, correct errors, eliminate redundant information, and expand the Bases to discuss the new requirements for steam generator blowdown isolation. These changes will have no effect on equipment operation. There will be no adverse effect on any design basis accident previously evaluated or on any equipment important to safety. Therefore, the proposed changes will not result in a significant increase in the probability or j                  consequences of an accident previously evaluated.
l The proposed changes have no adverse effect on any of the design basis accidents previously evaluated and have no adverse effect on how the RPS and ESFAS function to mitigate the consequences of design basis accidents.
Therefore, the license amendment request does not impact the probability of an

     .           U. S. Nucitar R:gulatory Commission B17190/Attrchment 2/Pcg3 4 accident previously evaluated nor does it involve a significant increase in the consequences 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 will not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. They do not alter the way any structure, system, or component functions and do not alter the manner in which the plant is operated. The proposed changes do not introduce any new failure modes. Therefore, the proposed                                        l changes will not create the possibility of a new or different kind of accident from                            }
any accident previously evaluated.                                                                             I
: 3.      Involve a significant reduction in a margin of safety.
The proposed changes will correct the maximum reactor power level specified; change RPS trip setpoints, allowable values, and bypass setpoints; change ESFAS trip setpoints, allowable values, and block setpoint changes; add a new Technical Specification and additional requirements associated with the automatic isolation of steam generator blowdown; and make various minor editorial and non-technical changes. There will be no adverse effect on equipment important to safety. The RPS and ESFAS will continue to function as designed to mitigate the consequences of design basis accidents. Therefore, there will be no significant reduction of the margin of safety as defined in the Bases for the Technical Specifications affected by the proposed changes.
The NRC has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (March 6, 1986, 51 FR 7751) of amendments that are considered not likely to involve an SHC. The minor editorial and non-technical changes proposed herein to correct spelling errors, correct a
               - capitalization error, add page amendment numbers, add the specific plant parameter (steam generator pressure) to use if an RPS or ESF function can be bypassed, change
               ' the value of the parameter (pressurizer pressure) used in action statements, add a "5" symbol, change "value" to "setpoint," and update the index are enveloped by example (i), a purely administrative change to Technica! Specifications. The change proposed f                herein to add requirements for automatic steam generator blowdown isolation is enveloped L                by ex' ample (ii), a change that constitutes an additional limitation, restriction, or control l                not presently included in the Technical Specifications. The other changes proposed L                herein are not enveloped by a specific example.
1 As described above, this License Amendment Request does not impact the probability of an' accident previously evaluated, does not involve a significant increase in the consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident from any accident previously evaluated, and does not L_ _   _._ _      __        _               _        _                      _     _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ . _ _ - . _ _ _

l
   ,   U. S. Nuciser Regulatory Commission L.   . B17190/ Attachment 2/Paga 5 result in a significant reduction in a margin of safety.                                     Therefore, NNECO has concluded that the proposed changes do not involve an SHC.
3 1
l
                                                                                                                         \
                                                                                                                         .}}