View source for B17519, Application for Amend to License DPR-65,revising Loss of Normal Feedwater (Lonf) Analyses to TS 2.2.1.Addl TS Bases Change to Floor Value for Thermal Margin Low Pressure Reactor Trip & Proposed FSAR Changes,Included

{{Adams
| number = ML20198P944
| issue date = 12/28/1998
| title = Application for Amend to License DPR-65,revising Loss of Normal Feedwater (Lonf) Analyses to TS 2.2.1.Addl TS Bases Change to Floor Value for Thermal Margin Low Pressure Reactor Trip & Proposed FSAR Changes,Included
| 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 = B17519, NUDOCS 9901070168
| package number = ML20198P949
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 22
| project = 
| stage = Request
}}

=Text=
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nopaeny na. (no== 1w. we=if-w 06sas Northeast I.         '
Nuclear Energy-                                         '
mu       Nuden Power Station
.                                                                                                                               Northeast Nudear Energy Cosepany
                                                                                                                              , P.O. Box 128 l
Taterford, CT 06385-0128 l                                                                                                                               (860) 447-1791 Faz (860) 44 6-4277
                                                                                                                                %e Northeast Utilities Systein DEC 2 81998
                                                                                                                                                                                \
f                                                                                                                                       Docket No. 50-336 B17519 I
Re: 10CFR50.90               l i
U.' Si Nuclear Regulatory Commission
;              ; Attention: Document Control Desk l-            iWashington,' DC 20555 Millstone Nuclear Power Station, Unit No. 2 i_                                                                                         Proposed Revision to Technical Specifications
!                                                                                                  Loss of Normal Feedwater Flow introduction l
L Pursuant to 10CFR50.90,. Northeast Nuclear Energy Company (NNECO) hereby l                proposes to amend Operating License DPR 65 by incorporating the attached proposed
: i.               changes into the Technical Specifications of Mill:Mone Unit No. 2. NNECO is proposing to change Technical Specification 2.2.1, " Limiting Safety System Settings - Reactor Trip Setpoints," and the associated Bases. These proposed changes are associated
[
               ..with the revised locs of normal feedwater (LONF) analyses. An additional Technical Specification Bases change to the floor value for the thermal margin low pressure reactor trip is also includedJ This proposed change is not related to the revised LONF analyses.
                                                                                                                                                                           /
NNECO also proposes to amend; Operating License DPR-65 by incorporating the                                                                                 /,
attached change to the Millstone Unit No. 2 Final Safety Analysis Report (FSAR). The proposed changes to the FSAR, except.the. floor value for the thermal margin low
               ; pressure reactor trip,'are associated with the revised LONF analyses.                                                                                     Mf
: Attachment 1 provides a discussion of the proposed changes and the Safety Suramary.
Attachment 2 provides the Significant Hazards Consideration. Attachment 3 provides 1the marked-up version of the appropriate pages of the current Technical Specifications.
                - Attachment 4 provides the retyped pages of the Technical Specifications. Attachment 5 provides the change to the Millstone Unit No. 2 FSAR.
s,-
9901070168 981228 DR :ADOCK 05000336
                .,                                                                        PDR om:2.s uus.es                                                                                                                           ,                             )
_w   _-_x --_        - - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ - _ _ _ _ - _ _                                                                      .
                                                                                                                                                                      .. .   .J

U.S. Nuclear Regul tory Commission B17519/Page 2 The proposed changes to Technical Specification 2.1.2, are on the same page, 2-4, which has been proposed to be changed in separate letters dated July 21,1998,W and November 10,1998.* The proposed change to Technical Specification Bases Page B 2-7 is on the same dated July 21,1998.                     Thegage        which proposed     hascontained changes been proposed in this letter to do be  changed in a separate not assume approval of any of the previously submitted changes.
Environmental Considerations 1
NNECO has reviewed the proposed License Amendment Request against the criteria i   of 10CFR51.22 for environmental considerations.                            The proposed Technical Specification changes will modify the low steam generator water level reactor trip i
setpoint and bases to be consistent with the revised LONF analyses, and modify the
;   bases for a reactor trip to account for instrument uncertainty. The proposed FSAR changes include credit fcc automatic auxiliary feedwater actuation and atmospheric steam dump valve operation. These changes do not significantly increase the type and amounts of effluents that may be released oif site. In addit lon, this amendment request will not significantly increase 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                                                                                                    {
The following items included in the proposed changes to the Technical Specifications and the FSAR analyses are required to meet the safety analysis limits.
: 1. An increase in the low steam generator water level reactor trip setpoint.
: 2.      Credit for automatic initiation of the motor driven auxiliary feedwater l
pumps.
: 3.      Credit for operation of the atmospheric steam dump valves.
In aggregate, these changes involve Unreviewed Safety ' :estions.
W     M. L. Bowling letter to the NRC, ' Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Features Trip Setpoints," dated July 21,1998.
A     M. L. Bowling letter to the NRC, " Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Gpecifications Reactor Prctective and Engineered Safety Feature Actuation System Instrumentation," dated November 10,1998.                                                                !
* M. L. Bowling letter to the NRC, " Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Features Trip Setpoints," dated July 21,1998.
                                                                             .                                   a

i             'U.S. Nucircr Regul tory Commission
      -        B17519/Page 3 The purpose of these changes is to prevent steam generator dryout following a LONF             .i '
event.' By preventing steam generator dryout, sufficient removal of decay heat from the -
l               Reactor Coolant = System (RCS) will occur, preventing excessive RCS heatup and
:               pressurization. This will ensure the steam generator fatigue analysis remains valid,
              ' and excessive discharge of primary coolant through the pressurizer safety valves does i               not occur. Therefore, there will be no adverse effect on the consequences of a LONF.
,               event.
The revised LONF analyses indicate that the appropriate acceptance criteria are met, j'              including no s'eam generator dryout. As a result, NNECO has concluded the proposed changes are safe. In addition, the proposed changes do not involve a significant l,              impact on pubib heatfn and safety (see the Safety Summary provided in Attachment 1)            ;
and do not involve a Significant Hazards Consideration pursuant to the provisions of
;              - 10CFR50.92 (see the Significant Hazards Consideration provided in Attachment 2).
j               Therefore, NNECO requests the NRC review and approve the proposed changes to the .           .:
Millstone Unit No.' 2 Technical Specifications and FSAR through an amendment to Operating License DPR-65, pursuant to 10CFR50.90.
Plant Ob :ations Review Committee and Nuclear Safety Assessment Board
:               The Plant Operations Review Committee and Nuclear Safety Assessment Board have reviewed and concurred with the determinations, i
:               Schedule 1
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.
N 4
4 5
4' 4                                                                                                              i

   ,. _ U.S. Nuclear Reguictory Commission 817519/Page 4 ff you should have any questions on the above, please contact Mr. Ravi Joshi at (860)                                                  ,
440-2080.
Very truly yours, a
NORTHEAST NUCLEAR ENERGY COMPANY i
M. L. Bowling, Jr.                  /
Recovery Officer - Technical Services -
4 i
         . Sworn to and subscribed before me
)         this $7ff day of [Er <e 8<#                                   1998                                                                     l i                                  Notary Public                                                                                                 >
l.0RETTA F. G000 SON                                                                        :
: l.        My Commission expires                         NOTARY PilRt 10                                                                           ,
Commission Expires November 30,2001
]                                                                                                                                                 l l
Attachments (5)                                                                                                                         ;
cc:       H. J. Miller, Region 1 Administrator S. Dembek, NRC Project Manager, Millstone Unit No. 2 D. P. Beaulieu, Senior Resident inspector, Millstone Unit No. 2                                                               .
W. M. Dean, Director, Millstone Project Directorate                                                                           l W. D. Lanning, Director, Millstone inspections                                                                              .i J. P. Durr, Chief, inspections Branch E. V. Imbro, Director, Millstone ICAVP inspections Director                                                                                                                      l Bureau of Air Management                                                                                                      l Monitoring and Radiation Division                                                                                             l Department of Environmental Protection                                                                                        !
79 Elm Street -
Hartford, CT 06106-5127 i
                                                                             . . _ _ _ _ _      _ _ - _ _ = _ _ _ _ _ _ _ _ _ _ - _ _ _ .

t ,i t
l Docket No. 50-336 B17519 l
l l
l I
I Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Loss of Normal Feedwater Flow l             Discussion of Proposed Changes i
l l
l 1
I l
December 1998
                                                                                                          )

U. S. Nucl:ar Regulitory Commission                                                                .
l
                     - B17519/ Attachment 1/Page 1 i
L                                            Proposed Revision to Technical Specifications i                                                          ' Loss of Normal Feedwater Flow
                                .                         - Discussion of Proposed Changes i
  ;                   introduction                                                                                       ,
a
)                                      .
j                    . Northeast Nuclear Energy Company (NNECO) hereby proposes to amend Operating j                     License DPR65 by incorporating the attached proposed changes into the Technical
                    . Specifications of. Millstone Unit No. 2. NNECO is proposing to change Technical i
Specification 2.2.1, " Limiting Safety System Settings - Reactor Trip Setpoints," and the j''                   associated Bases.' These proposed changes are associated with the revised loss of normal lfeedwater (LONF) analyses. An additional Technical Specification Bases
[                     change to the floor value for the thermal margin low pressure reactor trip is also included. This proposed change is not related to the revised LONF analyses.
I                     NNECO'also proposes to amend Operating License DPR-65 by incorporating the 1                     attached changes to the Millstone Unit No. 2 Final Safety Analysis Report (FSAR). The              i L                     changes to FSAR Chapters 10 and 14 are associated with the revised LONF analyses.
i                     The LONF analysis located in FSAR Chapter 10 is a best estimate analysis. It is l                     performed to demonstrate that the Auxiliary Feedwater (AFW) System meets the
!                     reliability requirement in that one of the three AFW pumps delivers sufficient flow to prevent steam generator dryout, and to remove decay heat. The LONF analysis                        ,
e                     located in FSAR Chapter 14 is the design basis accident analysis. It is performed to i                     verify that the protective systems at Millstone Unit No. 2 can successfully mitigate a l
LONF event.
The revised LONF analyses will correct a non-conservative steam generator liquid p                     inventory assumption, at the time of the reactor trip on low steam generator water level, that j                     is contained in the current LONF analyses. This was reported in Licensee Event Report
[
LER-98 012-00.* A reduction in the assumed auxiliary feedwater delivery rates, resulting from a revised hydraulic analysis of the AFW System, has also been incorporated.
i                     The revised FSAR Chapter 14 LONF analysis has determined that the current steam
!'                    generator water level low reactor trip setpoint, which is based on an analytical limit of
!                     34% narrow range level, is not acceptable. The analytical limit needs to be increased
!                     to 43% narrow range level to prevent steam generator dryout. If steam generator dryout occurs, sufficient decay heat will not be removed from the Reactor Coolant
;                     System (RCS) resulting in excessive RCS heatup and pressurization. By preventing steam generator dryout,-the steam generator fatigue analysis will remain valid, and                ,
!                    -excessive discharge of primary coolant through the pressurizer safety valves will not occur.
                    '
* Licensee Event Report, LER-98-012-00, " Millstone Nuclear Power Station Unit 2 - Nonconservative g                            Assumptions in the Facility Loss of Normal Feedwater Analysis," dated June 22,1998.
2
               \
s  y        4                         ,.L   - - -

3            U. S. Nuclur Rsgulttory Commission e-                B175;9/ Attachment 1/Page 2 in addition, the revised Chapter 14 LONF analysis has shown that neither the current                                                                                                                               i
:                 steam generator water level low reactor trip analytical setpoint (34.0%), or the proposed                                                                                                                          l
;                 analytical setroint (43%) provides sufficient water inventory in the steam generators at                                                                                                                           j i                the time of trip such that auxiliary feedwater will not be required for 10 minutes. This                                                                                                                          ;
:                 contradicts the current Technical Specification Basis for the steam generator water level i                 low reactor trip setpoint (Technical Specification 2.2.1).                                                                                                                                                         :
i                                                                                                                                                                                                                                    l i                The . current Chapter 14 LONF analysis _ assumes that the motor driven auxiliary feedwater (MDAFW) pumps become available following operator action with no credit                                                                                                                                  l l                for automatic initiation. The revised Chapter 14 analysis, in addition to increasing the
!                 analytical limit for the low steam generator water level reactor trip, assumes that the                                                                                                                           '
MDAFW pumps actuate automatically within 4 minutes after steam generator water
.                 level reaches the automatic auxiliary feedwater actuation setpoint. Automatic actuation
:                 of the MDAFW pumps, which is currently provided by the Auxiliary Feedwater
.                 Automatic initiation System (AFAIS), has been designed with sufficient redundancy to 4
assure that no single failure results in the loss of the protective function. However,                                                                                                                            1
:                -additional systems are credited in the revised analysis to assure acceptable l                 performance of the AFW System. Therefore, this change could result in an increase in
;                 the probability of occurrence of a malfunction of the MDAFW pumps.                                                                                                                                                j
                                                                                                                                                                                                                                    )
(
:                 The FSAR Chapter 10 LONF analysis (best estimate) has always credited automatic actuation of auxiliary feedwater to demonstrate that a single MDAFW pump is sufficient
,                 to remove decay heat. However, the revised analysis now credits operation of the                                                                                                                                  ]
steam generator atmospheric dump valves, instead of the steam generator safety                                                                                                                                     ;
i                 valves as in the current analysis, so that lower steam generator pressures, and                                                                                                                                    l
:-                subsequent higher auxiliary feedwater flow rates can be predicted. Without crediting
:                 the atmospheric dump valves, and the resultant higher auxiliary feedwater flow, steam generator dryout could occur resulting in an RCS heatup. This could result in opening                                                                                                                              ;
the pressurizer power operated relief valves or safety valves, increasing the probability                                                                                                                          i of occurrence of a malfunction of these valves.                                                                                                                                                                    l l
-                 The results of the revised LONF analyses, which are contained in FSAR Chapters 10 and 14, have concluded that the LONF event does not result in the violation of the                                                                                                                                 l F                 Specified Acceptable Fuel Design Limits, that the peak pressurizer pressure does not                                                                                                                               j j                 exceed 110% of the design pressure, that liquid primary coolant is not expelled through
;                 the pressurizer safety valves, and that adequate cooling water is supplied by the AFW                                                                                                                              l System to prevent steam generator dryout and allow a safe snd orderly plant shutdown.
i An additional change to the Millstone Unit No. 2 Technical Specification Bases, not l                 related to the revised LONF analyses, has been included in this License Amendment Request. This change to the thermal margin low pressure reactor trip floor value is                                                                                                                                j presented after the discussion of the revised LONF analyses,                                                                                                                                                       i J
J
   ~     v                                    -              - - - - - - - _ _ _ + _ _ _ . _ _ - - - - - . . - - - - - - _ _ _ _ . - _ - _ - _ . - - - - - _ - _ _ . - - . - _ . _ _ _ - _ _ . - - - _ - - - - - .

   ,   U. S. Nuclear Reguictory Commission
  -    B17519/ Attachment 1/Page 3 Backaround                                                                                                    ;
     ' The Technical Specification Basis for the low steam generator water level reactor trip, which has not changed since the issuance of the original Millstone Unit No. 2 Operating                       ;
License in 1975,* was based on manual operator action to initiate auxiliary fee Mater                          i flow ten minutes after a LONF evenL in response to the NUREG-0737, Millstone Unit                              l No. 2 made a design change to automatically initiate the two MDAFW                                           .l was added to Technical Specifications by License Amendment                               Even though     No. 63."
                                                                                                                      !  pum Technical Specification requirements for the automatic initiation of auxiliary feedwater                       l were added, the Technical Specification Basis for the low steam generator water level                          l reactor trip was not changed because the FSAR Chapter 14 LONF analysis of record                               l did not credit initiation of auxiliary feedwater prior to 10 minutes.
i The AFW System is designed to provide feedwater to the steam generators for the removal of sensible and decay heat from the Reactor Coolant System in the event of a                          '
      ~ loss of normal feedwater flow. The AFW System is used to mitigate transients and accidents such as Loss of Normal Feedwater, Small Break Loss of Coolant Accident, and Station Blackout. The system has two redundant electric motor driven pumps                                 l (each one can supply adequate feedwater espacity for decay heat removal) and one turbine driven pump (rated at twice the capacity of one electric motor driven pump). An automatic initiation signal starts both electric motor driven pumps. The turbine driven pump is manually started by a control room operator.
The current LONF analysis of record, provided in Section 14.2.7 of the Millstone Unit No. 2 FSAR, dates back to Cycle 10 when Siemens Power Corporation became the fuel vendor. At that time, Millstone Unit No. 2 still had the original steam generators installed. Consistent with the previous analysis, this analysis only credited auxiliary feedwater initiation ten minutes after the LONF event. At the end of Cycle 11, the
      . steam generators were replaced. The LONF event was not reanalyzed following replacement of the steam generators. It was determined that the full power steam generator liquid inventory in the replacement steam generators at the low water level reactor trip setpoint was only approximately 6400 pounds mass (ibm) less than the p
original steam generators at the low water level reactor trip setpoint. Considering this small reduction in the steam generator liquid inventory, and the available margin to i
steam generator dryout in the Cycle 10 LONF analysis, it was determined that the LONF event did not warrant reanalysis following the replacement of the steam gener,cors.
As a ' result of the Millstone Unit No. 2 review conducted in response to the NRC request pursuant to 10CFR50.54'f) dated April 16, 1997,") and the Independent Corrective Action Verification Program order. dated August 14, 1996,* the LONF
        % 0.' D. Parr letter to D. C. Switzer, issuance of Operating License No. DPR-65, dated August 1,1975.
       ? T. M. Novak letter to W. G. Counsil, Issuance of Amendment 63, dated January 14,1981.
        ") S. J. Collins letter to N. S. Cams, NRC request pursuant to 10CFR50.54(f), dated April 16,1997.
o      * " W. T. Russell letter to T. C. Feigenbaum, " Confirmatory order Establishing independent Corrective

_ U. S. Nucimr Regulttoy Commission B17519/ Attachment 1/Pege 4
!   analyses contained in FSAR Chapters 10 and 14 have been revised. The revised l   analyses include steam gerterator blowdown flow (no flow was modeled before), and a j   reduction in AFW flow delivery to the steam generators to provide margin to account for pump degradation. These changes, together with the reduced inventory of the i   replacement steam generators, challenge the acceptance criteria for a LONF event i   During the reanalysis of the LONF event with the replacement steam generators, it was i   determined that acceptable consequences (avoidance of steam generator dryout) could
:   not be demonstrated waiting 10 minutes for auxiliary feedwater flow. An increase in the analytical limit for the reactor trip on low steam generator water level and cradit for the i   automatic initiation of auxiliary feedwater are required to avoid steam generator dryout.
The reason that steam generator dryout would occur is because the steady state full power steam generator liquid inventory in the replacement steam generators is
:   approximately 19,000 lbm less than that of the original steam generators at the low j   water level reactor trip analysis setpoint of 34% narrow range level. Using a RELAP                               ,
. model for the reanalysis of the LONF event, Siemens Power Corporation has determined that the transient liquid inven%ry in the steam generators at the low water
-   level reactor trip setpoint is an additional 15,000 lbm less than the steady state inventory at the low water level reactor trip setpoint. This additional reduction in inventory appears to be a result of the redistribution of inventory in the steam generator i   due to the loss of subcooling in the downcomer region.                            The Siemens Power Corporation RELAP model was benchmarked to the steady stete steam generator
. inventories provided with the replacement steam generators. The Cycle 10 LONF
!   analysis was initialized at the steam generator low water level reactor trip setpoint, ana                        '
the steady state liquid inventory matched the data provided with the original steam j   generators.
Additional Technical Specification changes, not included in this License Amendment j    Request, will increase the setpoint for automatic auxiliary feedwater actuation on steam
;   generator water level to account for instrument uncertainty, and add automatic isolation of steam generator blowdown to conserve steam generator water inventory. This will occur at the same steam generator water level as automatic auxiliary feedwater                                   -
actuation. These changes have been submitted in a {{letter dated|date=July 21, 1998|text=letter dated July 21,1998}}.*
FSAR Chanaes The Millstone Unit No. 2 FSAR Chapter 14 LONF event has been revised such that the steam generator liquid inventory assumption at the time of reactor trip on low steam generator water level is consistent with the design of the replacement steam generators. -The revised LONF analysis also incorporates a reduction in auxiliary Action Verification Program (Effective immediately) - Millstone Nuclear Power Station, Units 1, 2 and 3," dated August 14,1996.
     * ' M. L. Bowling, Jr. letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2, Proposed Revision to Technical Specifications, Reactor Protection and Engineered Safety Features Trip Setpoints," dated July 21,1998.

  , U. S. Nucirr Regulttory Commission B17519/ Attachment 1/Page 5 feedwater delivery rates resulting from a recalculation of the AFW System flows. The results of this reanalysis have lead to the conclusion that the analytical limit for the low steam generator water level reactor trip must be raised to 43% narrow range level from the current 34% narrow range level. This will result in Technical Specification changes that will be discussed later. In addition, the revised LONF analysis will now take credit for automatic initiation of the MDAFW pumps within 4 minutes of reachir j the low steam generator water level automatic auxiliary feedwater actuaiion setpoint.
The Millstone Unit No. 2 FSAR Chapter 10 LONF analysis (best estimate) has also been revised to correct the steam generator liquid inventory assumption, consistent with the des gn of the replacement steam generators. The revised LONF analysis also incorporates a reduction in auxiliary feedwater delivery rates resulting from a recalculation of the AFW System flows. The assumption of automatic initiation of one MDAFW pump within 4 minutes, after the low steam generator water level auxiliary feedwater actuation setpoint is reached, has not changed. However, to demonstrate that one MDAFW pump delivers sufficient flow to preclude steam generator dryout, credit has been taken in the revised analysis for the operation of the steam generator atmospheric dump valves, instead of the main steam safety valves as in the current analysis. This new assumption yields lower predicted steam generator pressures, resulting in increased auxiliary feedwater flow.
The proposed changes to the Millstone Unit No. 2 FSAR associated with the revised LONF analyses are:
: 1. FSAR Section 7.2.3.3.4, which describes the steam generator low water level reactor trip setpoint, is being modified by deleting the statement that there is sufficient water inventory in the steam generators at the time c' ; rip to provide a margin of more than 13 minutes before auxiliary feedwater is required. A new statement is being added that there is sufficient water inventory in the steam generators at the time of trip to assure that steam generator dryout does not occur before auxiliary feedwater delivers sufficient flow to remove decay heat and recover steam generator water level.
: 2.      FSAR Section 10.4.5.3, Auxiliary Feedwater System, is being modified by changing the reference in the paragraph which states that a MDAFW pump is sufficient to remove decay heat. The new reference corresponds to the best estimate LONF analysis with the corrected replacement steam generator liquid inventory.
: 3.      FSAR Section 14.2.7 is being modified to reflect the assumptions and results of the revised LONF analysis, consistent with the replacement steam generators. It will now be assumed the reactor trips at a steam generator water level of 43%
narrow range (analytical limit), the MDAFW pumps start automatically within 4

U. S. Nucl:: r R:gul: tory Commission
          . B17519/ Attachment 1/Page 6 minutes of AFW System actuation on low steam generator water level, and the steam driven AFW pump becomes available due to operator action 10 minutes following the reactor trip on low steam generator water level.
: 4.      FSAR Section 14.2.7 is also being modified to incorporate additional editorial changes.
LONF Analysis Related Technical Specification Chanaes Changes to the Technical Specifications are necessary to ensure the validity of the revised analyses. These changes are discussed below.
Technical Specification 2.2.1 The trip setpoint and allowable value for the low steam generator water level reactor trip contained in Table 2.2-1, " Reactor Protective Instrumentation Trip Setpoint Limits,"
will be changed to be consistent with the revised LONF analyses. The revised analyses assume an analytical limit of 43% narrow range level, instead of the current analytical limit of 34% narrow range level. The calculation of the trip setpoint and allowable value, which includes instrument uncertainty, has determined that the trip setpoint should be changed from t 36.0% to 2 48.5%, and the allowable value should be changed from 2 35.2% to 2 47.5%.
The effsets of a harsh environment (pressure, temperature, and radiation) have not been included in the setpoint and allowable value calculations. The setpoint and                                          j' allovsable value calculation method is consistent with the approach used for a previous Technical Specification change that was submitted by a {{letter dated|date=July 21, 1998|text=letter dated July 21,1998}},*
and supplemented by a {{letter dated|date=October 6, 1998|text=letter dated October 6,1998}}.'')                                                                      l Technical Specification 2.2.1 Bases l
The basis for the steam generator water level low reactor trip will be modified to be
             . consistent with the revised Chapter 14 LONF analysis. The discussion concerning available water inventory and time until auxiliary feedwater is required will be removed.
The proposed change to the FSAR will include a discussion of the relationship between the LONF analysis and the need to automatically initiate auxiliary feedwater flow.
m M. L. Bowling, Jr. letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2, Proposed Revision to Technical Specifications. Reactor Protection and Engineered Safety Features Trip Setpoints," dated July 21,1998.
(*)  M. L. Bowling, Jr. letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, l
Unit No. 2, Response to a Request for Additional Information Regarding Tecnnical Specification Amendment Request Reactor Protection and Engineered Safety Features Trip Setpoints (TAC NO.
MA2340),* dated October 6,1998.
_  _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .        .- 1

U. G. Nucinr R:guintory Commission B17519/ Attachment 1/Page 7 Non LONF Analysis Related Technical Specification Bases and FSAR Chanae The following Technical Specification Bases and associated FSAR change is not related to the revised LONF analyses.
Technical Specification 2.2.1 Bases The basis for the thermal margin low pressure reactor trip will be modified. The current basis states that the floor, or minimum value, for this trip function is set at 1850 psia.
This value will be changed to be consistent with instrument uncertainty calculations that have determined that the floor should be increased to 1865 psia. The increase in floor value is the result of greater instrument uncertainties when harsh containment environment conditions are included. The calculation method used is consistent with the approach used for the proposed low steam generator water level reactor trip setpoint.
Safety Summary The analysis of a LONF event, as described in the Millstone Unit No. 2 FSAR Chapters 10 and 14, has been revised. Certain key assumptions have been changed to ensure acceptable analysis results. An evaluation of the safety impact of the LONF analyses changes, and associated Technical Specification changes will be presented.                In addition, an evaluation of the safety impact of an additional non LONF analysis related Technical Specification Bases change is included.
LONF Analyses Changes The LONF analyses, contained in FSAR Chapters 10 and 14, have been revised using a steam generator liquid inventory assumption, at the time of reactor trip on low steam generater water level, that is consistent with the oesign of the replacement steam generatas. The revised Chapter 10 and 14 LONF analyses also incorporate a reduction in auxiliary feedwater delivery rates resulting from a recalculation of the AFW System flows. The results of the revised analyses indicate that the analytical limit for the low steam generator water level reactor trip must be raised to 43% narrow range level from the current 34% narrow range level. This will result in a change to the low steam generator water level reactor trip setpoint listed in Technical Specification 2.2.1.
The revised Chapter 14 LONF analysis will now take credit for automatic initiation of the MDAFW pumps. The current Chapter 14 LONF analysis assumes auxiliary feedwater flow will be initiated 10 minutes after the event. The Chapter 10 LONF analysis assumption of automatic initiation of one MDAFW pump within 4 minutes, after the low steam generator water level automatic auxiliary feedwater actuation setpoint is reached, has not changed.
1 l

;          U. S. Nucl:cr Regulatory Commission B17519/ Attachment 1/Page 8                                                                           l l
l          To demonstrate th'at one MDAFW pump delivers sufficient flow to preclude steam                        >
generator dryout, the Chapter 10 LONF analysis will now take credit for the operation l          of the steam generator atmospheric dump valves, instead of the main steam safety i          valves as in the current analysis. This new assumption yields lower predicted steam l          generator pressures which result in an increase in the delivered AFW flow.
l l          The results of the revised Chapter 10 and 14 LONF analyses have concluded that the                    ,
i LONF event does not result in the violation of the Specified Acceptable Fuel Design                   ,
,          Limits, that the peak pressurizer pressure does not exceed 110% of the design
,          pressure, that liquid primary coolant is not expelled through the pressurizer safety valves, and that adequate cooling water is supplied by the AFW system to prevent steam generator dryout and allow a safe and orderly plant shutdown. Therefore, there
.          will be no adverse impact on public health and safety as a result of the revised
          ' analyses.
j          LONF Analyses Related Technical Specification Changes The trip setpoint and allowable value for the low steam generator water level reactor l-         trip will be changed to be consistent with the revised LONF analyses which assume an analytical limit of 43% narrow range level, instead of the current analytical limit of 34%
l          narrow range level. The calculation of the trip setpoint, which includes instrument uncertainty, has determined that the trip setpoint should be changed from E 36.0% to i         3 48.5 %.
!          The increase in the low steam generator water level Reactor Protection System (RPS)
:          actuation setpoint from E 36% to E 48.5% could result in an increase in the probability of an         ;
RPS actuation on low steam generator level since the difference between the proposed i           setpoint and the normal operating value of steam generator level will decrease. The i           proposed actuation setpoint is below the normal operating level of 60 to 75%. Steam                  .
j           generator water level is not expected to approach the actuation setpoint during normal
!           operation. An unexpected plant event (e.g., loss of main feedwater or difficulty controlling j           steam generator level at low power levels) would be necessary for steam generator level to
,           approach the actuation setpoint. To provide the operators with advance notice of the steam i           generator low level condition, the existing RPS low steam generator water level pretrip alarm setpoint will be changed to provide approximately the same margin between pretrip i           and trip as currently exists (5%). This will ensure that the pretrip alarm is received prior to i"          reaching the actual reactor trip setpoint. Since the probability of the need for an RPS actuation on low steam generator water level will increase, the probability of a malfunction of the RPS to actuate when called upon will increase. Therefore, the proposed change will result in an increase in the probability of occurrence of a previously evaluated malfunction of equipment important to safety. However, it is not expected that the proposed change will result in any significant impact on plant operation since steam generator water level is normally maintained well above the
: j.          proposed reactor trip setpoint. Even though the proposed change will decrease the
: j.         margin between the normal operating steam generator level and the RPS actuation I
l                                                                                                                l l

1 U. S. Nucl=r R::gulatory Commission
;-            B17519/ Attachment 1/Page 9
!             setpoint, this change will not significantly impact the probability of an RPS actuation on 4             low steam generator level during normal plant operations. Therefore, the proposed                     t j             change will not significantly increase the probability of occurrence of a malfunction of              -
the RPS to actuate.
The basis for the steam generator water level low reactor trip will be modified to be l           consistent with the revised LONF analysis. The discussion concerning available water                  i i
inventory and time until auxiliary feedwater is required will be removed. The proposed j             change to the FSAR will include a discussion of the relationship between the LONF                     :
l          . analysis and the need to automatically initiate auxiliary feedwater flow.                            l
;             The proposed change to the steam generator low water level reactor trip setpoint and 1
allowable value will ensure a reactor trip signal is generated at, or before the analytical
;              limit used in the LONF analyses is reached. This setpoint change will not adversely                  ,
j              affect RPS operation. The RPS will continue to function as before. Therefore, there will be no adverse impact on public health and safety.
Non LONF Analyses Related Technical Specification Bases and FSAR Change
!             This Technical Specification Bases and FSAR change is not related to the revised LONF analyses.
The basis for the thermal margin low pressure (TMLP) reactor trip (Technical                           l Specification 2.2.1 Bases) will be modified. The current basis states that the floor, or              i 4             minimum value, for this trip function is set at 1850 psia. This value will be changed to              l
[              be consistent with instrument uncertainty calculations that have determined that the                  l floor should be increased to 1865 psia. The increase in floor value is the result of                   l greater instrument uncertainties when harsh containment environment conditions are
,              included.
i-The increase in the TMLP floor (from 1850 psia to 1865 psia) could result in an increase l
 ;             in the probability of an RPS actuation on thermal margin low pressure since the difference
:              between the proposed floor setpoint and the normal operating value of pressurizer pressure i             will decrease. However, the proposed actuation setpoint is significantly below the normal operating pressure of approximately 2250 psia. Pressurizer pressure is not expected to
:              approach the actuation setpoint during normal operation. A significant plant event (e.g.,
j              loss of primary coolant) would be necessary for a rapid pressure excursion to approach the
             = actuation setpoint. Since the setpoint change is small, it will not adversely impact the probability of an RPS actuation on low pressurizer pressure during normal plant operations, and will not result in any significant restriction to plant operation. The RPS
: i.           will continue to function as before.
i f .
      - ..--   e r             ,      .x, ..,

  .   ._         ._ _ .._ _      _ _ . ..        - _ __._.~...__. _ _ . _ . _ _ _ _ - - _ , _
    ~
U. S. Nucle:r R2gulatory Commission
:*       B17519/ Attachment 1/Page 10 The proposed change has no effect on how the RPS functions to mitigate the e        consequences of design basis accidents. Also, the proposed change has no adverse l        offect on any design basis accident previously evaluated. Therefore, there is no
;        adverse impact on public health and safety.
i
!        Conclusion l        The proposed changes have no effect on how the RPS and AFW System function to mitigate the consequences of design basis accidents. Also, the proposed changes have no adverse effect on any design basis accident previously evaluated. Therefore, there is no adverse impact on public health and safety.
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i l
1
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1 Docket No. 50-336 i 817519 s
Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Loss of Normal Feedwater Flow Significant Hazards Consideration F
d December 1998

U. S. Nuclear RegulClory Commission 7      B17519/ Attachment 2/Page 1 Proposed Revision to Technical Specifications Loss of Normal Feedwater Flow Significant Hazards Consideration Sionsficant Hazards 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 analysis.of a loss of normal feedwater (LONF) event, as described in the Millstone Unit No. 2 FSAR Chapters 10 and 14, has been revised. Certain key assumptions have been changed to ensure acceptable analysis results. An                       )
evaluation of the LONF analyses changes, and associated Technical Specification' changes will be presented. In addition, an evaluation of an additional non LONF analyses related Technical Specification Bases and FSAR                   ;
                     ' change is included.
                                                                                                                   ]
                   ' LONF Analyses Changes
                   ' The LONF analyses, contained in FSAR Chapters 10 and 14, have been revised                    ,
using a steam generator liquid inventory assumption, at the time of reactor trip
:                    on low steam generator water level, that is consistent with the design of the replacement steam generators. The revised Chapter 10 and 14 LONF analyses j                    also incorporate a reduction in auxiliary feedwater delivery rates resulting from a recalculation of the Auxiliary Feedwater (AFW) System flows. The results of l                    revised analyses indicate that the analytical limit for the low steam generator t                    water level reactor trip must be raised to 43% narrow range level from the
;                    current 34% narrow range level. This will result in a change to the low steam generator water level reactor trip setpoint listed in Technical Specification 2.2.1.
l l                    The revised Chapter 14 LONF analysis will now take credit for automatic                       l initiation of the motor driven auxiliary feedwater (MDAFW) pumps. The current Chapter 14 LONF analysis assumes auxiliary feedwater flow will be initiated 10                ,
minutes after the event.                   The Chapter 10 LONF analysis assumption of
                    ' automatic initiation of one MDAFW pump within 4 minutes, after the low steam
.."                  generator level AFW actuation setpoint is reached, has not changed.
IL l-
                                           - . . ~ . .                         ,

J o.
l-          U. S. Nucitar Regulatory Commission
.-          B17519/ Attachment 2/Page 2
.                  To demonstrate that one MDAFW pump delivers sufficient flow to preclude
;                  steam generator dryout, the Chapter 10 LONF analysis will now take credit for              ,
:                 the operation of the steam generator atmospheric dump valves, instead of th_e               i i                  main steam safety valves as in the current analysis. This new assumption yields             l lower predicted steam generator pressures which result in an increase in the l
delivered AFW flows.
LONF Analyses Related Technical Specification Changes l
!                  The trip setpoint and allowable value for the low steam generator water level             l reactor trip will be changed to be consistent with the revised LONF analyses.
;                  The revised analyses assume an analytical limit of 43% narrow range level, instead of the current analytical limit of 34% narrow range level. The calculation l
[                  of the trip setpoint, which includes instrument uncertainty, has determined that          !
;-                 the trip setpoint should be changed from 2 36.0% to 2 48.5%.                              l l                  The increase in the low steam generator level Reactor Protection System (RPS)              :
l                  actuation setpoint from 136% to 148.5% will result in an increase in the probability
!                  of an RPS actuation on low steam generator water level since the difference between the proposed setpoint and the normal operating value of steam generator l
i                  level will decrease. The proposed actuation setpoint is below the normal operating        :
!                  level of 60 to 75%. Steam generator level is not expected to approach the actuation j                  setpoint during normal operation. An unexpected plant event (e.g., loss of main j                  feedwater or difficulty controlling steam generator level at low power levels) would l-                 be necessary for steam generator level to approach the actuation setpoint. To L
provide the operators with advance notice of the steam generator low level condition, the existing RPS low steam generator water level pretrip alarm setpoint will be changed to provide approximately the same margin between pretrip and trip as currently exists (5%). This will ensure that the pretrip alarm is received prior to
                  . reaching the actual reactor trip setpoint. Therefore, even though the proposed
(                  change will decrease the margin between the normal operating steam generator i                   level and the RPS actuation setpoint, this change will not significantly impact the      l l                   probability of an RPS actuation on low steam generator level during normal plant         !
: l.                  operations. In addition, the proposed setpoint and allowable value change will
;                   ensure a reactor trip signal is generated at, or before the analytical limit used in     j the revised LONF analyses is reached. Therefore, the RPS will continue to
.                  function as designed to mitigate the consequences of the design basis accidents.                                                                               ;
j                  The basis for the steam generator level low reactor trip will be modified to be l                   consistent with the revised LONF analysis. The discussion concerning available water inventory and time until auxiliary feedwater is required will be removed.           i
:                   The proposed change to the FSAR will include a discussion of the relationship between the LONF analysis and the need to automatically initiate auxiliary d

   .                                                                                               I U. S. Nucl:ar Regul tory Commission B17519/ Attachment 2/Page 3                                                                   !
feedwater flow.
Non LONF Analyses Related Technical Specification Bases and FSAR Change This Technical Specification Bases and FSAR change is not related to the revised LONF analyses.
The basis for the thermal margin low pressure (TMLP) reactor trip (Technical            l Specification 2.2.1 Bases) will be modified. The current basis states that the          I floor, or minimum value, for this trip function is set at 1850 psia. This value will
:           be changed to be consistent with instrument uncertainty calculations that have determined that the floor should be increased to 1865 psia. The increase in floor value is the result of greater instrument uncertainties when harsh containment environment conditions are included.
The increase in the TMLP floor (from 1850 psia to 1865 psia) could result in an increase in the probability of an RPS actuation on thermal margin low pressure since the difference between the proposed floor setpoint and the normal operating      ;
value of pressurizer pressure will decrease. However, the proposed actuation setpoint is significantly below the normal operating pressure of approximately 2250 4
psia. Pressurizer pressure is not expected to approach the actuation setpoint during normal operation. A significant plant event (e.g., loss of primary coolant) would be necessary for a rapid pressure excursion to approach the actuation setpoint. Since the setpoint change is small, it will not adversely impact the probability of an RPS actuation on low pressurizer pressure during normal plant operations. In addition the proposed change to the floor value will ensure a reactor trip signal is generated at, or before the analytical limit used in the respective accident analyses is reached. Therefore, the RPS will continue to function as designed to mitigate the consequences of the design basis accidents.
l Conclusion                                                                            l The results of the revised LONF analyses contained in FSAR Chapters 10 and            l 14 have concluded that the LONF event does not result in the violation of the Specified Acceptable Fuel Design Limits, that the peak pressurizer pressure           i does not exceed 110% of the design pressure, that liquid primary coolant is not expelled through the pressurizer safety valves, and that adequate cooling water is supplied by the AFW System to prevent steam generator dryout and allow a safe and orderly plant shutdown. By preventing steam generator dryout, sufficient removal of decay heat from the Reactor Coolant System (RCS) will occur, preventing excessive RCS heatup and pressurization. This will ensure the steam generator fatigue analysis remains valid, and excessive discharge of primary coolant through the pressurizer safety valves does not occur.
Therefore, there_ will be no adverse effect on the consequences of a LONF event. This is consistent with the acceptance criteria contained in Standard

t j           -                                                                                                                )
;'             U. S. NuclCar RCgulatory Commission
              . B17519/ Attachment 2/Page 4-Review Plan (SRP) 15.2.7.* (Millstone Unit No. 2 is not an SRP plant.)
~
The proposed changes do not alter the way any structure, system, or component i                       functions. The changes in actuation setpoints and equipment used in the LONF analyses affect equipment important to the mitigation of design basis accidents.
l These changes do not affect any equipment that can cause a design basis accident                    ,
t'                     to occur. Therefore, the proposed changes do not affect the probability of                          j occurrence of a previously evaluated accident.
t
!                       These proposed changes' do not alter tM way any structure, system, or
,                       component functions. There will be no adverse effect on any design basis accident previously evaluated, on any equipment important to safety, or on the
,                     ' radiological consequences of any design basis accident. Therefore, these                           '
proposed changes will not adversely affect the consequences of a previously l                       evaluated accident.
)              2.       Create the possibility of a new or different kind of accident from any accident previously evaluated.
{
Results of the proposed LONF analyses have demonstrated that the Specified 4
Acceptable Fuel Design Limits are not violated, that the peak pressurizer and
                      ' steam generator pressures do not exceed 110% of the design pressure, that                            l 4
liquid primary coolant is not expelled through the pressurizer safety valves, and                    j that adeouate cooling water is supplied by the AFW System to prevent steam                           i generator dryout and allow a safe and orderly plant shutdown. Therefore, there                       l are no new or different types of failures of systems or equipment important to
'                       safety which could cause a new or different type of accident from any accident                       i previously evaluated.                                                                               j
!                                                                                                                            I j                       The proposed changes will not alter the plant configuration (no new or different                     j i                       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                            )
changes will not create the possibility of a new or different kind of accident from                 l any accident previously evaluated.                                                                 j t
: 3.     . Involve a significant reduction in a margin of safety.
The revised FSAR Chapter 14 analysis has concluded that the steam generator I
low water level reactor trip setpoint does not provide sufficient water inventory in
                                                    ~
the steam generators at the time of the reactor trip such that auxiliary feedwater
;                    ' flow will not be required for 10 minutes. This contradicts the current Technical N        Standard Review Plan (SRP) 15.2.7, " Loss of Normal Feedwater Flow," Rev.1 - July 1981.
g__

U. S. Nucl:ar R:gul tory Commission B17519/ Attachment 2/Page 5 Specification Basis (Technical Specification 2.2.1) for the steam generator low water level reactor trip setpoint. Therefore, the revised analysis reduces the
.              margin of safety as defined in the Bases of the Millstone Unit No. 2 Technical Specifications. However, with the proposed changes to increase the low steam generator water level reactor trip setpoint and taking credit for automatic AFW System actuation, it has been shown that operation of these systems can mitigate the LONF event, and ensure plant response is within the acceptance criteria. Results of the proposed LONF analyses have demonstrated that the Specified Acceptable Fuel Design Limits are not violated, that the peak pressurizer and steam generator pressures do not exceed 110% of the design pressure, that liquid primary coolant is not expelled through the pressurizer safety valves, and that adequate cooling water is supplied by the AFW System to prevent steam generator dryout and allow a safe and orderly plant shutdown.
Therefore, these proposed changes do not involve a significant reduction in a margin of safety.
The proposed change to the floor value for the TMLP reactor trip function is the result of a revision to the instrument loop uncertainty and setpoint calculations.
The proposed change to the Technical Specification Basis will incorporate the RPS TMLP floor setpoint change. This change to the TMLP floor will not adversely affect this function. The TMLP reactor trip function will still operate as designed. The RPS will continue to function as designed to mitigate the consequences of design basis accidents. Therefore, this proposed change does not involve a significant reduction in a margin of safety.
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 changes proposed herein are not enveloped by any specific example.
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 result in a significant reduction in a margin of safety. Therefore, NNECO has concluded that the proposed changes do not involve an SHC.

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t Docket No. 50-336-B17519 i
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4 Attachment 3 Millstone Nuclear Power Station, Unit No. 2
                          - Proposed Revision to Technical Specifications Loss of Norrnal Feedwater Flow Marked Up Pages l
1 December 1998 1
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