View source for U-602196, Application for Amend to License NPF-62,requesting Changes to TS to Improve & Reformat CPS TS Consistent w/NUREG-1434, Improved BWR-6 TS, Rev 0,Sept 1992

{{Adams
| number = ML20062J296
| issue date = 10/26/1993
| title = Application for Amend to License NPF-62,requesting Changes to TS to Improve & Reformat CPS TS Consistent w/NUREG-1434, Improved BWR-6 TS, Rev 0,Sept 1992
| author name = Perry J
| author affiliation = ILLINOIS POWER CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000461
| license number = 
| contact person = 
| case reference number = RTR-NUREG-1434
| document report number = U-602196, NUDOCS 9311020179
| package number = ML20062J300
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 68
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:____________ _ __-____ - --__ - .                    . . . . . . -
Uhnois Power Company Cknton Power Stadon P.O Box 678 Chnton. IL 61727 Tel 217 935-c226 Fax 217 9244G32 J. Stephen Perry Senor Vice Presdant ILLIN9tS                                                                         u-cozi96 POWER                                                                     1A7-93(io-26)tr EE.100a October 26,1993 Docket No. 50-461                                                              10CFR50.90 Document Control Desk Nuclear Regulatory Commission Washington, D.C. 20555

==Subject:==
Clinton Power Station Proposed Amendment of Facility Operating License No. NPF-62

==Dear Sir:==

Pursuant to 10CFR50.90, Illinois Power (IP) hereby applies for amendment of Facility Operating License No. NPF-62, Appendix A - Technical Specifications, for Clinton Power Station (CPS). This request consists of proposed changes to improve and reformat the CPS Technical Specifications consistent with NUREG-1434, " Improved BWR-6 Technical Specifications", Revision 0, September 1992.
This submittal represents a significant effort on the part of the BWR-6/ Mark III community to prepare consistent license amendment applications. As the BWR-6 owners have previously discussed with Mr. Grimes and others ofyour staff, implementation of this request will result in a significant contribution to increased safety in operation of CPS. In addition, this request implements cost saving efrorts on the part of both the NRC and IP in that significant NRC resource savings can be realized through generic review of the application of NUREG-1434 to the BWR-6 plants. Further, implementation of this request reduces the likelihood of future Technical Specification changes for CPS by reducing IP's backlog ofindividual line item changes.
As part of the development of this request, IP has applied the criteria contained in the Final NRC Policy Statement on Technical Specification Improvements to the current CPS Technical Specifications utilizing BWR Owners' Group (BWROG) report NEDO-31466, "Technica! Specification Screening Criteria Applicatio. and Risk Assessment,"
(and Supplement 1) as incorporated in NUREG-1434. The results of the application of the screening criteria for CPS are contained in Enclosure 1 to this letter.                                   f e
1 Enclosure 2 to this letter is arranged by Technical Specification section. For each          hf section two attachments are provided. Attachment I contains a markup of the current                       I CPS Technical Specifications reflecting the changes necessary to adopt NUREG-1434 for                       ,
CPS. Each of the changes is ar. notated to identify the applicable justification for the g In addition, No Significant Hazards Consideration (NSHC) evaluations are f $g                         1 9311020179 931026                  i, PDR      ADOCK 05000461                                                                       }   (i g p                          PDR      ?
1

U-602196 Page 2 of 2 prosided for each of the changes. Generic NSHC evaluations have been prosided at the beginning of Enclosure 2 for the changes annotated as "A" (administrative); "R" (relocated as identified in Enclosure 1); "M" (technical change - more restrictive; and "LA", "LB",
and "LC" (generic technical change - less restrictive). The justifications and the NSilC evaluations for those changes annotated as "L" (specific technical change - less restrictive) are prosided at the end of the associated Attachment 1.
Attachment 2 to Enclosure 2 of this letter contains a markup of NUREG-1434 to reflect the changes necessary to apply the NUREG to CPS. Each of the changes are annotated to identify the reasons for the change. Similar to Attachment I to Enclosure 2, each of the changes has been categorized for review purposes. The changes are identified as "B" (prosides plant-specific information contained in brackets in NUREG-1434), "P" (provides plant specific information not contained in brackets in NUREG-1434), or "C" (denoting a generic change to NUREG-1434). The discussion / justification for each change to NUREG-1434 is contained at the end of the associated Attachment 2.
l l
An aflidavit supporting the facts set forth in this letter and its enclosures is        l provided in the attachment to this letter.
IP has reviewed the proposed changes against the criteria of 10CFR51.22 for             )
categorical exclasion from environmental impact considerations. The proposed changes            I do not involve a significant hazards consideration, or significantly increase the amounts or change the types of efiluents that may be released offsite, nor do they significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, IP concludes that the proposed changes meet the criteria given in 10CFR51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Statement.
In closing, this request represents expenditure of significant resources on the part of the BWR-6 utiliiest and culminates an accomplishment that sets a precedence in the industry. IP looks forward to working with your staffin obtaining your approval of this request.
Sincerely yours,
                                                                  \'u
                                                                    . Perry Senior Vice President DAS/nis Attachment Enclosures cc:     NRC Clinton Licensing Project Manager NRC Resident Office, V-690 Regional Administrator, Region III, USNRC Illinois Department ofNuclear Safety
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l

Machment to U-602196
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STATE OF ILLINOIS COUNTY OF DEWITT J. Stephen Perry, being first duly sworn, deposes and says: That he is Senior Vice President of Illinois Power Company; that the application for amendment of Facility Operating License NPF-62 has been prepared under his supenision and direction; that he knows the contents thereof; and that to the best v: his knowledge and belief said application and the facts contained therein are tme and correct.
DATE:           This   26       day of October 1993 Signed:                             _ v[              M
                                                                                        . Sthhhen Perry Subscribed and sworn to before me this 4//>                           day of October 1993.
(  /
1                               ene Notary Public lz
:                 "OmCIAL SEAL'                   f Linda S. French                 u 0       Notary Pub!ic, State ofIEnois               h g My Commiren Expires 9/1/96 )[

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APPLICATION OF SELECTION CRITERIA TO THE CLINTON POWER STATION TECHNICAL SPECIFICATIONS

Enclosure 1 to U-602196 Page 2 of 65 CONTENTS Pace
: 1. INTRODUCTION   . . . . . . . . . . . . . . . . . . . . .                                         1-1
: 2. SELECTION CRITERIA      . . . . . . . . . . . . . . . . . .                                     2-1
: 3. PROBABILISTIC RISK ASSESSMENT (PRA) INSIGHTS                                    . . . . .       3-1
: 4. RESULTS OF APPLICATION OF SELECTION CRITERIA                                    . . . . .       4-1 APPENDIX A. Justification for Specification Relocation

g Enclosure I to U-602196 Page 3 of 65
: 1. INTRODUCTION The purpose of this document is to confirm the results of the BWR Owners' Group (BWROG) application of the Technical Specification selection criteria on a plant specific basis for Clinton Power Station (CPS).                   Illinois Power Company (IP) has applied the selection criteria to each of the current CPS Technical Specifica-tions         utilizing           the   BWROG      report        NEDO-31466,    " Technical Specification Screening Criteria Application and Risk Assessment" (and Supplement 1), as incorporated in NUREG-1434, "BWR-6 Improved Technical Specifications," Revision O.                         Additionally, in accordance with the NRC guidance, this confirmation of the application of selection criteria to CPS includes confirming the risk insights from Probabilistic Risk Assessment (PRA) evaluations, provided in the NEDO documents, as applicable to CPS.
CLINTON POWER STATION                           1-1                           Introduction
                                                                     ----_                                                J

Enclosure I to U.602196 Page 4 of 65
: 2. SELECTION CRITERIA IP has utilized the selection criteria provided in the NRC Final Policy Statement on Technical Specification Improvements of July 16, 1993, to develop the results contained in the attached matrix.
Probabilistic Risk Assessment (PRA) insights as used in the BWROG submittal were utilized, confirmed by IP, and are discussed in the next section of this report. The selection criteria and discussion provided in the NRC Final Policy Statement are as follows:
Criterion 1:      Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary:
Discussion of Criterion 1:       A basic concept in the adequate protection of the public health and safety is the prevention of accidents.        Instrumentation is installed to detect significant abnormal degradation of the reactor coolant pressure boundary so as to allow operator actions to either correct the condition or to shut down the plant safely, thus reducing the likelihood of a loss-of-coolant accident. This criterion is intended to ensure that Technical Specifications control those instruments specifically installed to detect excessive reactor coolant system leakage.
Criterion 2:      A process variable that is an initial condition of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier:
CLINTON POWER STATION                 2-1                           Selection Critoria
(                                                  .

Enclosure 1 to U-602196 Page 5 of 65 Discussion of Criterion 2:        Another basic concept in the adequate protection of the public health and safety is that the plant shall be operated within the bounds of the initial conditions assumed in the existing Design Basis Accident and Transient Analyses.      These analyses consist of postulated events, analyzed in the Updated Safety Analysis Report (USAR) ,
for which    a  structure,    system,   or    component must meet specified functional goals.      These analyses are contained in Chapters 6 and 15 of the USAR (or equivalent chapters) and are identified as Condition II, III, or IV events (ANSI N18.2) (or equivalent) that either assume the failure of or present a challenge to the integrity of a fission product barrier.
As used in Criterion     2,  process variables are only those parameters for which specific values or ranges of values have been chosen as reference boureds in the Design Basis Accident or Transient Analyses and shAch are monitored and controlled during power operation such that process values remain within the analysis bounds.
The purpose of this criterion is to capture those process variables that have initial values assumed in the Design Basis Accident and Transient Analyses, and which are monitored and controlled during power operation. So long as these variables are maintained within the established values, risk to the public safety is presumed to be acceptably low.
Criterion 3:   A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier:
CLINTON POWER STATION             2-2                 Selection Criteria

Enclosure 1 to U-6021%
Page 6 of 65 Discussion of Criterion 3:      A third concept in the adequate protection of the public health and safety is that in the event that a postulated Design Basis Accident or Transient should occur,     structures,    systems,             and components are available to function or to actuate in order to. mitigate the consequences of the Design Basis Accident or Transient.
Safety sequence analyses (or their equivalent) have been performed in recent years and provide a method of presenting the plant response to an accident.               These can be used to define the primary success paths.
A safety sequence analysis is a systematic examination of the actions required to mitigate the consequences of events considered in the plant's Design Basis Accident and Transient Analyses, as presented in Chapters 6 and 15 of the plant's Updated Safety Analysis Report (or equivalent chapters) . Such a safety sequence analysis considers all applicable events, whether explicitly or implicitly presented.                                         The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate                                                                       ,
(including consideration of the single failure criteria),                                                    so that the    plant response to Design Basis Accidents                                                        and Transients limits the consequences of these events to within the appropriate acceptance criteria.
It is the intent of this criterion to capture into Technical Specifications only those structures, systems, and components that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function.
Criterion 4:    A structure, system, or component which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety:
CLINTON POWER STATION              2-3                                   Selection Criteria

Enclosure I to U-602196 Page 7 of 65 Discussion of Criterion 4:    It is the Commission's policy that licensees retain in their Technical Specifications LCOs, action statements and Surveillance Requirements for the following systems (as applicable) which operating experience and PSA have generally shown to be significant to public health and safety, as well as any other structures, systems, or components that meet this criterion:
* Reactor    Core    Isolation    Cooling                      (RCIC)/ Isolation Condenser;
* Residual Heat Removal (RHR);
* Standby Liquid Control (SLC) ; and
* Recirculation Pump Trip (RPT).
The Commission recognizes that other structures, systems, or components may meet this criterion.           Plant- and design-specific PSAs have yielded valuable insight to unique plant vulnerabilities not fully recognized in the safety analysis report Design Basis Accident or Transient analyses.                                               It is the intent of this criterion that those requirements which PSA or operating experience exposes as significant to public health and safety-consistent with the Commission's Safety Goal and Severe Accident Policies-be retained or included in Technical Specifications.
CLINTON POWER STATION                             2-4              Selection Criteria

Enclosure I to U-602196 Page 8 of 65 The Commission expects that licensees, in preparing their Technical Specifications-related submittals, will utilize any plant specific PSA or risk survey and any available literature on risk insights and PSAs. This material should be employed to strengthen the technical bases for those requirements that remain in Technical Specifications, when applicable, and to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk. Similarly, the NRC staff will also employ risk insights and PSAs in evaluating Technical Specifications-related submittals. Furthermore, as part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to better use risk and reliability information to define future generic Technical Specification requirements.
CLINTON POWER STATION           2-5                                         Selection Criteria

. - - - - - . - - . -                -   -       .   -. .-      - . .      .~.--      .  .   . . - . . . _ . - -            -- - -- ._
4                   P Enclosure 1 to U-6021%
Page 9 of 65
: 3. PROBABILISTIC RISK ASSESSMENT (PRA) INSIGHTS Introduction and Obiectives The Final Policy Statement includes a statement that the NRC expects Owners Groups to utilize the available literature on risk insights to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk.
Those Technical Specifications proposed as being relocated to other plant controlled documents will be maintained ' under the 10 CFR I
50.59 review program. These Specifications have been compared to a variety of PRA material with two purposes: (1) to identify if a component or variable is addressed by PRA, and 2) if addressed, to                                               ,
judge if the component or variable is risk-important.              In addition,                                  )
in some cases risk was judged independent of any specific PRA material.      The intent of the review was to provide a supplemental                                             I screen to the deterministic criteria.              Technical Specifications to l
be retained were not reviewed.                This review was accomplished .in                                   l BWROG submittal NEDO-31466 and Supplement-1, except where discussed in Appendix A,      '' Justification For Specification Relocation," and has been confirmed by IP for those Specifications to be relocated.
Assumptions and Approach Briefly, the approach used in NEDO-31466 and Supplement 1 was the following-                                                                                                       1 l
The risk assessment analysis evaluated the loss of function of the system or component whose LCO was being considered for relocation and qualitatively assessed the associated effect on core damage frequency and offsite releases.               The assessment was based on available literature on plant risk insights and PRAs. The table provided at the end of this section lists the CLINTON POWER STATION                  3-1                         PRA Insights
,                                            -     -       --,                   - -                 -           -   -----.-,~.-m      ,

l Enclosure I to U-602196   .
Page 10 of 65 !
PRAs used to provide insights for making the assessments.       A                 ;
detailed quantitative calculation of the core damage and offsite release effects was not performed.        However, the analysis   did   provide  an   indication   of   the   relative                   l i
significance of those LCOs proposed for relocation on the                         ]
l likelihood or severity of the accident sequences that are I
commonly found to dominate plant safety risks. LCOs which did not meet the screening criteria were evaluated.           Those satisfying a criterion were not. The following analysis steps were performed for each LCO proposed for relocation:
: a. List the function (s) affected by removal of the LCO item.
: b. Determine the effect of loss of the LCO item on the function (s).
l
: c. Identify compensating provisions, redundancy, and backups                    j related to the loss of the LCO item.                                         l l
: d. Determine the relative probability (high, medium, or low) of the loss of the function (s) assuming the LCO item is removed from Technical Specifications and controlled by                      ,
I other procedures or programs.       Use information from current PRAs and related analyses to establish the relative probability.
: e. Determine the relative significance (high, medium, or low) of the loss of the function (s). Use information                    i from current PRAs and related analyses to establish the relative significance.                                                       I CLINTON POWER STATION           3-2                     PRA Insights

Enclosure I to U-602196 Page 11 of 65
: f. Apply risk category criteria to establish the potential risk significance or non-significance of the LCO item.
Risk categories were defined as follows:
RISK CRITERIA Consecuence Freauency        Hiah       Medium        Low                                              ,
High               S           S           NS
^
Medium             S           S           NS Low               NS          NS           NS                                              l l
S   =   Potential Significant Risk Contributor NS =    Risk Non-Significant l
: g. List any comments or caveats that apply to the above assessment.                                                                               l The output from the above evaluation was a list of LCOs proposed for relocation that could have potential plant safety risk   significance   if   not  properly    controlled  by   other procedures or programs.      As a result, these Specifications                                  ;
will be relocated to other plant controlled documents outside the Technical Specifications.
d l
i CLINTON POWER STATION             3-3                     PRA Insights l
1

Enclosure I to U-602196 Page 12 of 65 BWR PRAs USED IN NEDO-31466 RISK ASSESSMENT
* BWR/6 Standard Plant, GESSAR II, 238 Nuclear Island, BWR/6 Standard Plant Probabilistic Risk Assessment, Docket No. STN 50-447, March 1982.
* La Salle County Station, NEDO-31085,         Probabilistic Safety Analysis, Revision 1, February 1986.
* Grand Gulf Nuclear Station, IDCOR, Technical Report 86.2GG, Verification of IPE for Grand Gulf, March 1987.
* Limerick, Docket Nos. 50-352, 50-353, 1981, "Probabilistic Risk Assessment, Limerick Generating Station," Philadelphia Electric Company.
* Shoreham, Probabilistic Risk Assessment Shoreham Nuclear Power Station, Long Island Lighting Company, SAI-372-83-PA-01, June 24, 1983.
* Peach Bottom 2, NUREG-75/0104, " Reactor Safety Study," WASH-1400, October 1975.
* Millstone   Point   1,    NUREG/CR-3085,    " Interim   Reliability Evaluation Program:     Analysis of the Millstone Point Unit 1 Nuclear Power Plant," January 1983.
* Grand Gulf, NUREG/CR-1659, " Reactor Safety Study Methodology Applications Program: Grand Gulf #1 BWR Power Plant," October 1981.
* NEDC-30936P,   "BWR    Owners'    Group  Technical    Specification Improvement Methodology      (with Demonstration for BWR ECCS Activation Instrumentation) Part      2," June 1987.
CLINTON POWER STATION              3-4                      PRA Insights

                                                                                            'l l
l
   . .                                                                                         l Enclosure I to U-602196    j Page 13 of 65    j
: 4. RESULTS OF APPLICATION OF SELECTION CRITERIA The selection criteria from Section 2 were applied to the CPS Technical Specifications. The following is a summary of that application indicating which Specifications are being retained or relocated. Discussions that document the rationale for the relocation of each Specification which failed to meet the selection criteria are provided in Appendix A. 10 CFR 50.92 evaluations for those Specifications relocated are provided with the Discussion of Changes for the specific Technical Specifications.         IP will relocate those Specifications identified as not satisfying the criteria in a dedicated section of the CPS USAR,          programs,                     l I
procedures, or other licensee controlled documents.
a l
1 CLINTON POWER STATION           4-1         Results of Application

==SUMMARY==
DISPOSITION MATRIX Retained / Criterion old TS                   Title                                              New Ntrber                      for inclusion         Bases for Inclusion /Esclusion L,0    DEFINIT!ONS                                                           1.1                                  Yes             See Note 1 and Note 4.
3.10.2 3.10.3 3.10.4 M      SAFETY LIMITS                                                         M 2.1.1  THERMAL POWER, Low Pressure or Low Flow                               2.1.1.1                              Yes               See Note 2.
2.1.2  THERMAL POWER, High Pressure and High Flow                            2.1.1.2                              Yes              See Note 2.
2.1.3  Reactor Coolant System Pressure                                       2.1.2                                Yes              See Note 2.
2.1.4  Reactor vessel Water level                                            2.1.1.3                              Yes               See Note 2.
J 2      LIMtilNG SAFETY SYSTEM SETTING 2.2.1  Reactor Protection System (RPS) Instrtsnentation                      3.3.1.1                              Yes               The app (ication of Technical Specification selection criteria Setpoints                                                                                                                    is not appropriate. However, the RPS LSSS have been included as part of the RPS Instrwientation Specification, dich has been retained since the Functions either actuate to mitigate consequences of design basis accidents and transients or are retained as directed by the NRC as the Functions are part of the RPS.
J 3      LIMITING CONDITIONS FOR OPERAtt04 - APPLICABillfY 3.0.1  operational Conditions                                                LCO 3.0.1                            Yes               See Note 3.
3.0.2  Nonconpliance                                                         LCO 3.0.2                            Yes               See Note 3.
3.0.3  Generic Actions                                                       LCO 3.0.3                            Yes               See Note 3.
3.0.4  Entry into Operational Conditions                                     LCO 3.0.4                            Yes               See Note 3.
M      SURVEILLANCE REQUIREMENTS - APPLICABILITY 6.0.1  operational Conditions                                                SR 3.0.1                             Yes               See Note 3.
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                         - . _ __           m.       ._ __             . . ~ .. . . . .                   . _ . . _ _ _ ._.                        .                                       _. . _ _ _            .      .
                                                                                                                                                                                                                                              .t St##4ARY DISPOSITION MATRIX l
Retained / Criterion Old TS                                Title                                    New Nwber         for inclusion                     Bases for inclusion / Exclusion 4.0.2           Time of Performance                                             SR 3.0.2               Yes                         See Note 3.
l            4.0.3           Noncor9pt lance                                                 SR 3.0.3               Yes                         See Note 3.
4.0.4           Entry into Operational Conditions                               SR 3.0.4               Yes                         See Note 3.
4.0.5           ASME Code Class 1, 2, 3 Components                              5.7.10                 Yes                         See Note 3.
5.7.11 3/4,1           REACTIVITY CO4 TROL SYSTEM                                      3d Not a measured process variable, but is imortant parameter
           '3/4.1.1          Shutdown Margin                                                 3.1.1                  Yes-2 that is used to confirm the acceptability of the accident analysis.
3/4.1.2         Reactivity Anomalies                                            3.1.2                  Yes-2                       Confirms assmptions made in the reload saf-ty analysis.
3/4.1.3         Control Rods 3/6.1.3.1       Control Rod OPERABILITY                                         3.1.3                  Yes-3                       Primary success path in mitigating the consequences of 3.1.8                                              design basis accidents and transients.
3/4.1.3.2       control Rod Maximtsri scram Insertion Times                     3.1.3                  Yes-3                       Same as above.
;                                                                                            3.1.4 3/4.1.3.3       control Rod Scram Acetrutators                                  3.1.5                  Yes-3                       Same as above.
l                                                                                            3.9.5
           . 3/4.1.3.4       Control Rod Drive Coupting                                      3.1.'4                 Yes-3                       Same as above.
3/4.1.3.5       Control Rod Position Indication                                 3.1.3                  Yes-3                       Same as above.                                                                                  I 3.9.4 I
3/4.t.3.6       Control Rod Drive Hous;ng Support                               Relocated              No                          See Appendix A.
1, 3/4.1.4         Control Rod Program controls 3/4.1.4.1       Control Rod Withdrawat                                          3.3.2.1                Yes-3                       Prevents withdrawal of control rods that might exceed rod withdrawat error transient analysis asstaptions.
3.3.2.1                Yes-3                       Prevents withdrawal of out-of-sequence control rods that might                                'I
!            3/4.1.4.2       Rod Pattern Control System 3.1.3                                              set up high rod worth conditions beyond CRDA asstanptions. Also 3.1.6                                              prevents deviation beyond banked position withdrawal sequence                                   i that if vlotated could allow high rod worth conditions that l
would challenge the MCPR Safety Limit and 1 percent cladding                                    !
plastic strain fuel design limit during a rod withdrawat error                              @
event.                                                                                      P   ,
l 3/4.1.5         Stan & y Liquid Control System                                  3.1.7                  Yes-4                       Retained in accordance with the NRC Final Policy Statement on Technical Specification tinprovements due to potential risk f'
d
                                                                                                                                                                                                                                            ~
!                                                                                                                                               significance.                                                                                   ,
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* C:
CLIOTON POWER STATION                                                                    Page 2 of 16                                                                                      Stanary Disposition Matrix          y

==SUMMARY==
DISPOSITION MATRIX Retained / Criterion Old TS                                                                          Title                             New Ntsnber       for inclusion    Bases for inclusion / Exclusion 3/4.2                                                         POWER DISTRIBUTION LIMITS                            J 3
3/4.2.1                                                       Average Planer Linear Heat Generation Rate (APLMGR)  3.2.1                  Yes-2      Peak cladding tenperature following a LOCA is primarily dependent on initial APLNGR.         As such, it is an initlet condition of a DBA analysis.
3/4.2.2                                                        Deleted in Amendnent No.18 3/6.2.3                                                       Minfrun Critical Power Ratio                         3.2.2                  Yes-2      Utilized as an initial condition of the design basis analyses.
DBA analysis are performed to establish the largest reduction in Critical Power Ratio.      This value is added to the fuel cladding integrity safety limit to determine the MCPR value.
3/4.2.4                                                        Linear Heat Generation Rate (LMGR)                   3.2.3                  Yes-2      LNGR is calculated to avoid exceeding plastic strain timits on fuel rods. As such, it is an initial condition of Design Basis Transient Analyses.
W                                                              INSTRLHENTATION                                      L3 3/6.3.1                                                        Reactor Protection System Instrumentation            3.3.1.1 3/4.3.1.1                                                      Intermediate Range Monitors                          3.3.1.1                Yes        Retained as directed by the NRC, as it is part of the RPS System.
3/4.3.1.2                                                      Average Power Range Monitors                         3.3.1.1                Yes-3      Actuates to mitigate consequences of design basis accident or transient.
3/4.3.1.3                                                       Reactor Vessel Steam Dome Pressure-Nigh              3.3.1.1                Yes        Retained as directed by the NRC, as it is part of the RPS System.
3/4.3.1.4                                                       Reactor Vessel Water Level-tow, Level 3              3.3.1.1                Yes-3      Actuates to mitigate consequences of design basis accident or transient.
3/6.3.1.5                                                       Reactor Vessel Water Level-Nigh, Level 8             3.3.1.1                Yes-3      Actuates to mitigate consequences of design basis accident or transient.
3/4.3.1.6                                                      Main Steam Line Isolation Valve-Closure              3.3.1.1                Yes        Retained as directed by the NRC, as it is part of the RPS System.
3/4.3.1.7                                                      Main Steam Line Radiation-High                      Deleted                 No         Deleted; see RPS technical change for MSLRM.
3/4.3.1.8                                                      Drywell Pressure-High                               3.3.1.1                 Yes        Retained as directed by the NRC, as it is part of the RPS System.
3/4.3.1.9                                                       Scram Discharge Volume Water Levet-Nigh             3.3.1.1                 Yes        Retained as directed by the NRC, as it is part of the RPS System.                                                                                  m 3/6.3.1.10                                                      Turbine Stop Valve-closure                          3.3.1.1                 Yes-3      Actuates to mitigate consequences of design basis accident or                            $
transient,                                                                               g c::
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CLINTON POWER STATION                                                                                                          Page 3 of 16                                                    Sunnery Disposition Matrix

==SUMMARY==
DISPCstTION MATRIX l
Old is                         title                                                Retained / Criterion New Nunber        for Inclusion          Bases for Inclusion / Exclusion 3/4.3.1.11    Turbine Control valve Fast closure, Valve Trip system Dil Pressure-Low                       3.3.1.1                Yes-3 Actuates to mitigate consequences of design basis accident 3/4.3.1.12                                                                                                     or tra sient.
Reactor Mode switch shutdown Position                    3.3.1.1                Yes 3/4.3.1.13    Manual scram                                                                                     Retaind es directed by the NRC, as it is part of the RPs system.
3.3.1.1                 Yes 3/4.3.2        Containment and Reactor vessel isolation                                                         Retained as directed by the NRC, as it is part of the RPS system.
Control system                                          3.3.6.1 3.3.6.2 3/4.3.2.1     Primary and secondary contairvnent Isolation             3.3.6.1                 Yes-3 3.3.6.2                                 Actuates to mitigste the consequences of a DBA LOCA.
3/4.3.2.1.m   r4ain steam Line Radiation-High                          Deleted                No Deleted; see RPs technical change for MsLRM.
3/4.3.2.1.n   Manual Initiation                                        3.3.6.1                Yes 3.3.6.2                                 Retained system.
as directed by the NRC, as it is part of the Isolation 3/4.3.2.2     Main steam Line Isolation                                3.3.6.1                Yes-3' 3/4.3.2.2.b   Main steam Line Radiation-High                                                                   Actuates to mitigate the consequences of a psA LOCA.
Deleted                 No 3/4.3.2.2.1    Manual Initiation                                                                               Deleted; see RPs technical change for MSLRM.
3.3.6.1                 Yes Retained as directed by the NRC, as it is part of the Isolation system.
3/4.3.2.3      Reactor Water Cleanup system Isolation                  3.3.6.1                Yes-3' Actuates to isolate potential teakage paths to secondary 3/4.3.2.3.h   sLCs Initiation                                                                                  contairvnent consistent with safety analysis assumptions.
3.3.6.1                Yes-4 Retained due to i m ortance of sLCS and in accordance with the 3/4.3.2.3.1                                                                                                    NRC Final Policy statement on Technical specification Imrove-Manual Initiation                                        3.3.6.1 ments.
Yes Retained as directed by the NRC, as it is part of the Isolation system.
3/4.3.2.4     Reactor Core Isolation Cooling system Isolation          3.3.6.1                Yes-3' Actuates to isolate potentist teakage paths to secondary 3/4.3.2.4.c    RCIC steam supply Pressured ow                                                                  contairvnent consistent with safety analysis assumtions.
3.3.6.1                Yes Does not satisfy the selection criteria; however, it is being 3/4.3.2.4.d     RCIC Turbine Exhaust Diaphragm Pressure-High retained due to potentist risk significance.
3.3.6.1                Yes Does not satisfy the selection criteria; however, it is being 3/4.3.2.4.k    Manual Initiation                                                                                retained due to potential risk significance.
3.3.6.1                Yes Retained as directed by the NRC, as it is part of the Isolation system.
Except Anblent Temeratut'e and Differential Te Terature Instrunents, which are to be Relocated. see Appendix A.
P.,.
o CLINTON PouER STATION Page 4 of 16                                                                                       =w=  o surtnery Disposition Matrix     b "f  o l
RP lh h                                                                                                                                                                      -

==SUMMARY==
DISPOSITION MATRIX Retained / Criterion Old TS                                  Title                          New Nueer                           for inclusion                                            Bases for Inclusion / Exclusion 3/4.3.2.5        RHR System Isolation                                     3.3.6.1                                Yes-3'                                             Actuates to isolate potential leakage paths to secondary containment consistent with safety analysis assu mtions.
3/4.3.2.5.c      Reactor Vessel Water Level-tow, level 3                  3.3.6.1                                Yes                                                Retained due to importance of RHR System and the NRC Final Policy Statement on Technical Specification I mrovements.
3/4.3.2.5.e      Reactor Vessel (RNR Cut-in Permissive)                   3.3.6.1                                Yes-4                                              Retained due to igortance of the RMR System and in accorden.:e Pressure-4tigh                                                                                                                                     with the NRC Final Policy Statement on Technical Specification I@rovements.
3/4.3.2.5.g      Manual Initiation                                        3.3.6.1                                Yes                                                Retained as required by the NRC, as it is part of the Primary Containment Isolation System.
3/4.3.3          Emergency Core Croting System Actuation                  3.3.5.1 Instrumentation                                          3.3.8.1 3/4.3.3.A        Division    1 Trip System                                3.3.5.1                                Yes-3                                              ECCS mitigate the consequences of a DBA LOCA.
3/4.3.3.A.1.g Manual Initiation                                           3.3.5.1                                Yes                                                Retained as required by the NRC, as it is part of the ECCS Actuation System, 3/4.3.3.A.2.h Manual Inhibit ADS Switch                                   Relocated                              No                                                 See Appendix A.
3/4.3.3.A.2.f Manual Initiation                                           3.3.5.1                                Yes                                                Retained as required by the NRC, as it is part of the ECCS Attuation System.
3/4.3.3.B        Division II Trip System                                  3.3.5.1                                Yes-3                                              ECCS mitigate the consequences of a DBA LOCA.
3/4.3.3.8.1.g Manual Initiation                                          3.3.5.1                                 Yes'                                               Retained as required by the NRC, as it is part of the ECCS Actuation System.
3/4.3.3.B.2.g Manual Inhibit ADS Switch                                   Relocated                              No                                                 See Appendix A.
3/4.3.3.B.2.h Manual Initiation                                          3.3.5.1                                 Yes                                                Retained as required by the NRC, as it is part of the ECCS Actuatlon System.
j 3/4.3.3.C       Division III Trip System                                  3.3.5.1                                Yes-3                                              ECCS mitigate the consequences of a DBA LOCA.
3/4.3.3.C.1.c Reactor Vessel Water Level-4tigh, Level 8                   3.3.5.1                                Yes                                                Does not satisfy the selection criteria; however, it is being retained due to potential risk significence.
i 3/4.3.3.C.1.h Manual Initiation                                           3.3.5.1                                Yes                                                Retained as required by the NRC, as it is part of the ECCS Actuation System.
3/6.3.3.D        Loss of Power                                            3.3.8.1                                Yes-3                                               Loss of power instrumentation actuates to assure power availablLity to the ECCS in the event of a loss of of fsite h
power. Mitigation of DBAs relies on the availability of the ECCS and ECCS power supply.                                                     S
  *Except Asient Teversture and Dif ferentist Temperature Instruments, which are to be Relocated. See Appendix A.
8 5i ac t
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CLINTON POWER STATION                                                                              Page 5 of 16                                                                                                        Sumery Disposition Matrixo"o g .)
O 9 "5 .
l

==SUMMARY==
DISPOSITION MATRIX Retained / Criterion Old TS                            Title                            New Nunber       for inclusion                    Bases for inclusion / Exclusion 3/6.3.4        Recirculation Ptry Trip Actuation Instrumentation 3/4.3.4.1      ATuS Recirculation Ptry Trip System Instrumentation  3.3.4.2                                Yes-4     Retained due to inportance of ATWS Recirculation Ptrp Trip System and in accordance with the NRC Final Policy Statement on Technical Specification leprovements.
3/6.3.4.2      End-of Cycle Recirculation Ptry Trip System          3.3.4.1                               Yes-3      EOC RPT sids the reactor scram in protecting fuel cladding Instrtsnentation                                                                                      integrity by ensuring the fuel cladding integrity safety limit is not exceeded during a load rejection or turbine trip transient.
3/6.3.5        Reactor Core Isolation Cooting (RCIC) System         3.3.5.2                               Yes-3&4    Required to mitigate the consequences of a DBA and e atained due Actuation Instrtsnentation                                                                            to inportance of the RCIC System and in accordance with the NRC Final Policy Statement on Technical Specification Inprovements.
3/4.3.5.b      Reactor vesset Water Levet-High. Level 8             3.3.5.2                                Yes       Does not satisfy the selection criteria; however, it is being retained due to potentist rimk significance.
3/4.3.5.e      Manual Initiation                                    3.3.5.2                               Yes        Retained as required by the NRC, as it is part of the kCIC System Actuation System.
3/4.3.6        Control Rod Block Instrtsnentation                   3.3.2.1 3/4.3.6.1      Rod Pattern Control System                           3.3.2.1                               Yes-3      Prevents withdrawal of out-of sequence control rods that might set up high rod worth conditions beyond CRDA asstrptions. Also prevents devietion beyond a banked posttion withdrawat sequence that if vlotated could allow high rod worth conditions that would chattenge the MCPR Safety Limit and 1 percent cladding plastic strain fuel design limit during a rod withdrawat error event.
3/4.3.6.2      APRM                                                 Relocated                             No         See Appendix A.
3/4.3.G.3      Source Range Monitors                                Relocated                             No         See Appendix A.
3/6.3.6.4      Intertnediate Range Monitors                         Relocated                             No         See Appendix A.
3/4.3.6.5      Scram Discharge vottre                               Relocated                             No         See Appendix A.
3/4.3.6.6      Reactor Coolant System Recirculation Flow            Relocated                             No         See Agxandix A.
3/4.3.6.7      Rear: tor Mode Switch                                3.3.2.1                               Yes-3      Prevents control rods f rom being withdrawn when shutdown, thus 3.9.2                                            ensuring that inadvertent rod withdrawat error cannot occur during MODES 3 and 4.                                                             Ensures one-rod-out interlock is enforced,                                                                                                              y D '
n
                                                                                                                                                                                                                                        -e CLlafoN FoutR STATION                                                       Page 6 of 16                                                                                                                     Stsernary Disposition Matrix 5 h P, !2 68
                                                                                     - _ _ - _ _ _ . - _ - _ _ _ _ _                     - _ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ - _ _ _ _ .                             .___-____a

SUMwARY DISP 0SITION MATRIX Retained / Criterion Old TS                          Title                        New Ntrtier       for Inclusion    Bases for inclusion /Emetusion 3/4.3.7       Monitoring Instrunentation 3/4.3.7.1     Radiation Monitoring Instrtsnentation           3.3.7.1 3/4.3.7.1.1   Main Control Room Air Intake Radiation Monitor  3.3.7.1                Yes-3      Actuates to maintain control room habitability so that 3/4.3.7.1.2   Area Monitors                                   Relocated              No         See Appendix A.
3/4.3.7.2     Seismic Monitoring Instrts,entation             Relocated             No          See Appendix A.
3/6.3.7.3     Meteorological Monitoring Instrumentation       Relocated              No         See Appendix A.
3/6.3.7.4     Remote Shutdown Mo*11toring Instrinnentation    3.3.3.2                Yes        Does not satisfy the selection criteria; however, it is being retained as directed by the NRC as a significant contributor to risk reduction.
3/4.3.7.5     Accident Monitoring Instrtanentation            3.3.3.1                Yes-3      See Appendix A.
3/4.3.7.6     Source Range Monitors                           3.3.1.2               Yes         Does not satisfy the selection criteria; however, it is being retained because the NRC considers it necessary for flux monitoring during shutdown, startup and refueling operations.
3/4.3.7.7     Traversing In-Core Probe System                 Relocated             No          See Appendix A.
3/4.3.7.8     Chlorine Detection System                       Relocated             No          See Appendix A.
3/4.3.7.9     Removed by Previous Amer *ent 3/4.3.7.10    Loose-Part Detection System                     Relocated             No          See Appendix A.
3/4.3.7.11    Nain Condenser Offges Treatment System          Relocated             No          See Appendix A.
Explosive Gas Monitoring Instrumentation 3/4.3.8       Removed in Amentbent No. 60.
3/4.3.9       Plant Systems Actuation Instrumentation         3.3.6.3 3.3.6.4 3/4.3.9.1     Containment Spray System                        3.3.6.3               Yes-3       Actuates to mitigate consequences of DBA LOCA.
3/4.3.9.1.e   Manual Initiation                               3.3.6.3               Yes         Retained as required by the NRC, as it is part of the Containa ment Spray System Actuation System.
3/4.3.9.2     Feedwater System / Main Turbine Trip System     Relocated             No          See A @endix A.
3/4.3.9.3     Suppression Pool Makeup System                  3.3.6.4               Yes-3       Actuates to mitigate consequences of DBA LOCA.
3/4.3.9.3.e   SPMS Manual Initiation                          3.3.6.4                Yes        Retained as required by the NRC, as it is part of the SPMS System Actuation System.
3/6.3.9.3.f   SPMS Mode Switch Permissive                     Relocated             No          See Appendix A.
3/4.3.10      Nuclear Systems Protection Systee-              Relocated             No          See Appendix A.                                                      p.
Self Test System                                                                                                                                       g a
ac
* AE CLINTON POWER STATIDN                                                  Page 7 of 16                                                  Surinary Disposition Matrix h C
N

                                                                                                                                                                                                                                                                                                                                ~
i i

==SUMMARY==
DISPOSITION MATRIX I, '
Retained / Criterion Old TS                             Title                                                              New Ntrber       for Inclusion                                       Bases for inclusion / Exclusion                                                                                                   ,
i
                            }/_42 4,       REACTOR COOLANT SYSTEM                                                                  J 3                                                                                                                                                                                                    ,
3/4.6.1       Recirculation System                                                                                                                                                                                                                                                                        fl Recirculation Loops                                                                     3.4.1                                Yes-2                          Recirculation loop flow is an initlet condition in the safety                                                                   i 3/4.4.1.1                                                                                                                                                                  analysis. Closure of the flow control valves within specified                                                                        i 3.4.2 3.4.11                                                              time limits functions to mitigate the consequences of a LOCA.
t 3.4.3                                Yes-2                          Jet pupp OPERABILITY is asstaned in the LOCA analyses to assure 3/4.4.1.2      Jet Ptrps adequate core reflood capability.                                                                                                    i Recirculation Loop Flow                                                                3.4.1                                Yes 2                          Recirculation loop flow mismatch within timits is an inittet 3/4.6.1.3                                                                                                                                                                 condition in the safety analysis.
3.4.11                               Yes-2                          Tenperature dif ferentist between the reactor coolant in the                                                                         e 3/4.4.1.4      Idte Recirculation Loop startup reactor vessel and the idle loop is an initlet condition in the transient analysis.                                                           Idle loop startup with teeperatures outside the timit could result in a reactivity transient and potentist vlotation of the Safety Limit MCPR.
3/4.4.2        Safety Valves Safety / Relief valves                                                                 3.3.6.5                              Yes-3                          A minimin ruter of Safety /Retief Valves is assteed in the                                                                           -
3/4.4.2.1                                                                                                                                                                 safety analyses to mitigate overpressure events.
3.4.4 Safety / Relief Valves Low-Low Set Function                                            3.3.6.5                              Yes-3                           A mininun nteber of Safety / Relief valves is assumed in the                                                                        t 3/6.4.. 2                                                                                             3.6.1.6                                                              containment toeding safety analysis.                                                                                                ;
3/6.6.3        Reactor Coolant System Leakage Leakage Detection Systems                                                              3.4.7                                Tes-1                          Leak detection is used to indicate a significant abnormat 3/4.4.3.1                                                                                                                                                                  condition of the reactor coolant pressure boundary.
3.4.5                                Yes-2                           Leakage beyond timits would indicate a significant abnormal
!                             3/6.6.3.2     Operational Leakage                                                                                                                                         condition of the reactor coolant pressure boundary. Operation j                                                                                                                                  3.4.6 in this condition Is unanalyzed and may result in reactor                                                                           ,
coolant pressure boundary fatture.
Chemistry                                                                              Relocated                            No                              See Appendix A.
3/6.6.4 3
Specifle Activity                                                                      3.4.8                                Yes-2                           Specific activity provides an indication of the onset of I                              3/4.4.5                                                                                                                                                                  significant fuel cladding f ailure enri is a variable used in the DBA analysis.
3/4.6.6      Reactor Coolant System                                                                                                                                                                                                                                                                 O E
t                                                                                                                                                                                                                                                                                                                              *U "
50
* C i
Uh Stenary Disposition Matrin k tf j                              CLIOTON POWER STAil0N Page 8 of 16 5
08 i
_ _ - ___ - _ _.__ - __ - -                                            _ _ _         _ _ - - - . - ~ - _. ..                                                             .._ - - . - -- - . -... - - . .. - .-                                                                              _ . _- _ _    ,_-    _-_ --__          _

m

==SUMMARY==
DISPOSITION MATRIX Retained / Criterion Old TS                          Title          New Nurnber          for Inclusion    Bases for Inclusion / Exclusion 3/4.4.6.1     Pressure /fenperature Limits         3.4.11                  Yes-2     This LCO establishes inittat conditions for operation such that 5.8.1.7                            operation is prohibited in areas or at tenverature rate changes that might cause undetected flaws to propagate, in turn challenging the reactor coolant pressure boundary integrity.
3/4.4.6.2     Reactor Steam Dome                  3.4.12                   Yes-2     The reactor steam dome pressure is an initial condition for the overpressurization analyses.
3/4.4.7       Main Steam Line Isolation valves    3.6.1.3                  Yes-3     Main steam line isolation within specified time limits ensures that the retesse to the envirorvnent is consistent with the assmptions in the LOCA analysis.
3/4.4.8       Structurat Integrity                Relocated                No        See Appendix A.
3/4.4.9       Residual Heat Removal 3/4.4.9.1     Hot Shutdown                        3.4.9                    Yes-4     Retained in accordance with the NRC Final Policy Statement on 3.10.1                             Technical Specification Inprovements due to potential risk significance.
3/4.4.9.2     Cold Shutdown                       3.4.10                   Yes-4     Same as above 3.10.1 3/4.5         EMERGENCY CORE COOLING SYSTEMS      M S/4.5.1       ECCS-Operating                      3.5.1                    Yes-3     Functions to mitigate the consequences of a DBA.
5.8.2 3/4.5.2       ECCS-Shutdown                       3.5.2                    Yes-3     Functions to mitigate the consequences of a vesset draindown event.
3/4.5.3       Sumression Poot                     3.5.2                    Yes-2&3   Functions to mitigate the consequences of a DBA and a vessel 3.6.2.2                            draindown event.
3/4.6         CONIAINMENT SYSTEMS                 3d 3/4.6.1       Primary Contairunent 3/4.6.1.1     Primary Conteirinent Integrity      3.6.1.1                  Yes-2&3   containment integrity functions to mitigate the consequences 3.6.1.3                            of a DBA.
3/4.6.1.2     Primary Contalrunent Leakage        3.6.1.1                  Yes       this LCO does not satisfy the selectinn criterla; however,           R 3.6.1.2                            containment tenkage is an asstrption utilized in the LOCA           c 3.6.1.3                            safety analysis (but it is not a process variable). Therefore, it is being retained as a Surveillance Requirement.
{n
                                                                                                                                                       ?
                                                                                                                                                       %C
                                                                                                                                                       *C CLINTON POWER STATION                                      Page 9 of 16                                                    Stenary Disposition Matrix d 6 O k 55

SUMMAR ' DISPOSITION MATRIX Retained / Criterion Old TS                           Title                           New Ntrber         for inclusion      Bases for inclusion / Exclusion 3/4.6.1.3      Primary Contairvnent Air Locks                     3.6.1.2                Yes-3        Credit for air tightness is considered in safety analysis to limit offsite dose rates during a DBA.
3/4.6.1.4      MSiv Leakage control System                        3.6.1.8                Yes-3        Assumed in primary contairvnent isolation events to direct the release of untreated leakage from the MSIVs such that of fsite
                                                                                ~
dose is within 10 CFR 100 guidelines.
3/4.6.1.5       Containment Structural Integrity                   3.6.1.1                Yes-3        Containment functions to mitigate the conseg>ences of a DBA.
3/4.6.1.6       Containment Internal Pressure                      3.6.1.4                Yes-2        Containment pressure is an initial condition in the LOCA safety analysis.
3/4.6.1.7      Primary Containment Average Air Terperature         3.6.1.5                Yes-2        Primary containment air tecperature is an initial condition in the LOCA safety analysis.
3/4.G.1.8      Containmen* Building Ventilation and Purge Systems 3.6.1.3                 Yes-3        System is part of primary success path for en accident involving release of radioactivity offsite.
3/4.G.2        Drywell 3/4.6.2.1      Drywell Integrity                                  3.6.5.1                 Yes-2&3      Drywell integrity functions to mitigate the consequences of a 3.6.5.3                              DBA.
3/6.G.2.2      Drywell Bypass Leakage                             3.6.5.1                 Yes          This LCO does not satisfy the selection criteria; however, drywell bypass teskage is an asstrption utilized in the LOCA safety analysis (but it is not a process variable). Therefore, it is being retained as a Surveillance Regairement.
3/4.6.2.3      Drywell Air Locks                                  3.6.5.2                 Yes-3        Credit for drywell air lock teskage is an asstsption utilized in the LOCA safety analysis.
3/4.6.2.4      Drywell Structural Integrity                       3.6.5.1                 Yes-3        Drywell functions to mitigate the consequences of a DBA.
3/4.6.2.5      Drywell Internal Pressure                          3.6.5.4                Yes-2         Drywell pressure is an initlet condition in the LOCA safety analysis.
3/4.6.2.6      Drywet t Average Air Tecperature                   3.6.5.5                Yes-2         Drywell air terperature is an initial condition in the LOCA safety analysis.
3/4.6.2.7      Drywell Vent and Purge System                      3.6.5.3                Yes-3         Isolation valves function to limit consequences of a DBA LOCA.
3/4.6.3        Depressurlastion Systems 3/4.G.3.1      Suppression Pool                                   3.6.2.1                Yes-2&3       Suppression pool water level and te merature are initial 3.6.2.2                              conditions in the DBA LOCA analysis and mitigate the conse-quences of the DBA.
f 3/6.6.3.2      Containment spray                                  3.6.1.7                Yes-3         Contairenent spray is asstsned to mitigate the consequences of      3*
a DBA LOCA.                                                         E 8
m" AE ac bJ g:
CLIN 70N POWER STATION                                                     Page 10 of 16                                                      Struury Disposition Matrix y g 55

i

==SUMMARY==
DISPOSIY10N MATRIX i
Retained / Criterion Old TS                          Title                      New Ntsnber        for inclusion                            Bases for Inclusion / Exclusion 3/4.6.3.3     Suppression Pool Cooling                      3.6.2.3                 Yes-3                              Surpression pool cooling functions to limit the effects of a                                                                 ;
DBA.
3/4.G.3.4     Suppression Poot Makeup System                3.6.2.4                 Yes-3                              SPMU System functions to mitigate the consequences of a DBA LOCA.
3/4.6.4       Primary Containment Isolation valves          3.6.1.3                 Yes-3                              Isolation valves function to limit DBA consequences.
3/4.6.5      Drywell Post +LOCA Vacuum Relief Valves        3.6.5.6                 Yes-3                              Metps ensure drywell functions properly to mitigate the consequences of a DBA LOCA.
3/4.G.6      Secondary Contairnent 3/4.6.6.1     Secondary Containment Integrity               3.6.4.1                 Yes-3                              Secondary contairnent integrity is relied on ti, I'. nit the 3.6.4.2                                                    offsi:e dose during an accident by ensuring that any release to contairnent is delayed and treated before being released to the envirornent.
3/4.6.6.2    secondary Containment Automatic                3.6.4.2                 Yes-3                              Valve operation within time timits establishes secondary Isolation Denpers                                                                                        containment and limits of fsite dose releases to acceptable values.                                                                                                                      '
3/4.G.6.3    Standby Gas Treatment System                   3.6.4.3                 Yes-3                              Operation following a DBA acts to mitigate the consequences of 5.7.12                                                     offsite releases.
3/4.6.7      Atmosphere Control 3/4.6.7.1    Containment Hydrogen Recombiner Systems        3.6.3.1                 Yes-3                              Operates post LOCA to limit hydrogen and oxygen concentrations to below explosive concentrations that might otherwise challenge containment integrity.
3/6.G.7.2    Containment /Drywell Hydrogen Mixing System    3.6.3.3                 Yes-3                              operates post LOCA to limit hydrogen and oxygen concentrations to below explosive concentrations that might otherwise chattenge contaltnent integrity.
3/4.G.7.3    Primary Containment /Drywell Hydrogen         3.6.3.2                  Yes                                White this Specification does not meet any selection criteria, Ignition System .                                                                                         It is being retained as directed by the NRC. Operates post LOCA to timit hydrogen and oxygen concentrations to below explosive concentrations that might otherwise challenge containment integrity.
  . }f_42        PLANT SYSTEMS                                 M 3/4.F.1 3/4.P.1.1 Service Water Systems Shutdown Service Water System (Loops A, 8, C) 3.7.1                    Yes-3                              Designed for heat removat for safety related systems fattowing                                                           .
(
                                                                                                                                                                                                                                                      ~
<                                                              3.7.2                                                       a DBA. As such, acts to mitigate the consequences of an
[                                                                                                                          accident.
me 9
! .CLINTON POWER STAY!ON                                                 Page 11 of 16                                                                            Sm nary Disposition Metrix [ O4 Es

                 . _ -.                       --                                             .-= .     ._.      ~-. __
Staa8ARY DISPOSIT104 MATRlx Retained / Criterion old TS                            Title                                                    New Number          for inclusion                                       Bases for Inclusion / Exclusion 3/4.7.1.2       Ultimate Heat Sink                                                          3.7.1                    Yes-3                                         Heat sink for heat removal from safety related systems following a DBA. As such, acts to mitigate the consequences of an accident.
3/4.7.2         Control Roore ventitation System                                            3.7.3                    Yes-3                                         Mitigates the consequences of an accident by maintaining 3.7.4                                                                  habitability of the control room so that operators can remain 5.7.12                                                                 in the control room fottowing an accident and continue accident mitigation activities theref rom. It also functions to mitigate the consequences of an accident by ensuring that control room temerature is maintained such that control room safety reth*ed equipment remains OPERABLE following an accident.
3/4.7.3         Reactor Core Isolation Cooling System                                       3.5.3                    Yes-3&4                                       Required to mitigate the consequences of a DBA and retained in 3.3.5.2                                                                accordance with the NRC Finat Policy Statement on Technical Specification Igrovements due to potential risk significance.
3/4.7.4         Snubbers                                                                    Relocated                No                                            See Appendix A.
3/4. 7.5        Sealed Source Contamination                                                 Relocated                No                                            See Appendix A.
3/4.7.6         Main Turbine Bypass System                                                  3.7.6                    Yes-3                                         Actuates to mitigatw the consequences of a feedwater controtter f ailure-eaximin demand transient and a turbine trip with bypass event.
3/4.7.7         Liquid Storage Tanks                                                        5.7.13                   Yes                                           Does not satisfy selection criteria; however, it is retained as a program in Adninistrative Controts as directed by the NRC.
3/4.7.8         Main Condenser Offgas Monitoring 3/4.7.8.1       Offgas-Explosive Gas Mixture                                                s 7.13                   Yes                                           Does not satisfy selection criteria; however, it is retained as a program in Adninistrative Controls as directed by the NRC.
3/4.7.8.2       offges-Noble cas Radioactivity Rate                                         3.7.5                    Yes-2                                         Main condenser offges activity is an initial condition in the offjes system fatture event.
3/4.9           ELECTRICAL POWER SYSTEMS                                                    M 3/4.Q.1         AC Sources 3/4.8.1.1       AC Sources-operating                                                        3.8.1                    Yes-3                                         Required to mitigate the consequences of a L3A.
3.8.3 5.7.14 5.8.2 3/4.G.1.2       AC Sources-Shutdown                                                         3.8.2                    Yes-3                                         Functions to mitigate the consequences of a vessel dreirswn             E 3.8.3                                                                  event and is needed to support NRC Final Policy Statement               P-.
5.7.14                                                                 requirement for decay heat removat.                                     y 5.8.2                                                                                                                                          g m~
E5 ac CLINTON POWER STATION                                                                                Page 12 of 16                                                                                               Sumery Disposition Matrix d 9., t2 e :e

==SUMMARY==
DISPOSITl04 MAtRlx Retained / Criterion old TS                         Title                                                                                                                New Ntster                 for inclusion     Bases for inclusion / Exclusion 3/4.G.2     DC Sources 3/6.0.2.1    DC ">urces-operating                                                                                                                      3.8.4                          Yes-3      Required to mitigate the consequences of a DBA.
3.8.6 3/4.G.2.2    DC Sources-Shutdown                                                                                                                       3.8.5                          Yes-3      Functions to mitigate the consequences of a vessel draindown 3.8.6                                     event and is being retained to support the NRC Final Policy Statement requirement for decay heat remov31.
3/4.G.3      Onsite Power Distribution Systems 3/4.G.3.1    Distribution-Operating                                                                                                                    3.8.7                          Yes-3      Required to mitigate the consequences of a DBA.
3.8.9 3/4.G.3.2     Distribution-Shutdown                                                                                                                     3.8.8                          Yes-3      Functions to mitigate the consequences of a vessel draindown 3.8.10                                    event and is being retained to scoort the NRC Finst Policy Statement requirement for decay heat removal.
3/4.G.4       Electrical Equipment Protective Devices 3/4.0.4.1     Containment Penetration Conductor                                                                                                         Relocated                      No         See Appendix A.
Ovrrcurrent Protective Devices 3/4.0.4.2     Motor Operated Valves Thermal Overload Protection                                                                                         Relocated                      No         See Appendix A.
3/4.0.4.3     Reactor Protection System (RPS) Electric Power                                                                                            3.3.8.2                        Yes-3      Providas protection for the RPS bus powered instrtsnentation Monitoring                                                                                                                                                                          against unacceptable voltage and frequency conditions that could degrade the instrumentation so that it would not perferm          ,
the intended safety function.
* 3/4,9         REFUEL!NG OPERATIONS                                                                                                                      3_d 3/4.9.1       Reactor Mode Switch                                                                                                                      3.9.1                           Yes-3      Provides an interlock to preclude fuel loading with control 3.9.2                                      rods withdrawn. Operation is assumed in the control rod removal error during refueling and fuel assentity insertion error during refueling accident analysis.
3/4.9.2       Instrtsnentation                                                                                                                         3.3.1.2                         Yes        Does not satisfy selection criteria; however, it is retained because the NRC considers it necessary for flux monitoring during shutdown, startup and refueling ope' lons.                    [
o 3/4.9.3       Control Rod Position                                                                                                                     3.9.3                           Yes-3      All control rods must be fully inserted when loading fuel.           5*
This requirement is asstaned as an initial condition in the fuel     E assembly insertion error during refueling accident enslysis.         o 2
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Sumary Disposition Matrix * &
                                                                                                                                                                                                                                                                                                                 "C CLINTON POuER STATION                                                                                                                                                   Page 13 of *.6                                                                                 o kd E8

   . .             . _ . . _ _ _ _ _ _ .           .m. m m ._ _ . _ . . . _ . . .         _           m_.._                      m. __ ~ . . _ _ . _ . ._                      . .           .   -- .._ .<.. _ _ _.._ _ _ ..                                                      _ _ . -
i SUMMRY DISPOSITION MAT'IX Retained / Criterion old TS                                         Title                                                                            New Nurrber              for Inclusion            Bases for inclusion / Exclusion 3/4.9.4           Decay Time                                                                                                      Relocated                    No                 Although this LCO satisfied Criterion 2,                     the activities necessary prior to commencing movement of irradiated fuel ensure that there will always be 24 hours of suberiticality                                              :
before movement of any irradiated fuel.               Hence this Spe.ci-                                 r fication has been relocated.
3/6.9.5           Communications                                                                                                  Relocated                    No                 See Appendix A.
J       3/4.9.6            Fuel Handling Equioment 3/o.9.6.1         Refueling Platform                                                                                              Relocated                    No                 See Appendix A.
3/4.9.6.2         Auxiliary Platform                                                                                              Relocated                    No                 See Appendix A.
3/4.9.7           Crane Travel-Spent Fuel Storage *                                     .  *per                                   Relocated                    No                 See Appendix A.
Containment Fuel Pint, and New                           4- 1 St6              le Vault 3/4.9.8           tlater Level-4teactor Vessel                                                                                    3.9.6                        Yes-2              A minimum amount of water is recpired to assure adequate 3.9.7                                           scrubi,Ing of fission products following a fuel handling accident.
3/4.9.9           Water Level-Spent Fuel Storage and                                                                              3.7.7                        Yes-2              Same as above.
i Upper Containment Fuel Poots 3/4.9.10          control Rod Removal l       3/4.9.10.1        Single control Rod Removal                                                                                      3.10.4                       Yes                See Note 4 j                                                                                                                                         3.10.5 t
3/4.9.10.2        Multiple Control Rod Removat                                                                                    3.10.6                       Yes                See Note 4 3/4.9.11          Residual Heat Removal and Coolant Circulation 3/4.9.11.1        High Water Level                                                                                                3.9.8                        Yes-4              Retained in accordance with the NRC Final Policy Statement on Tes. .ical Specification Inprovements due to potential risk significance.
3/6.9.11.2        Low Water Level                                                                                                 3.9.9                        Yes-4              Same as above.
j     - 3/4.9.12           Inclined Fuel Transfer System                                                                                  Relocated                    No                 See Appendix A.
3/4.10            SPECIAL TEST EXCEPTIONS                                                                                         3,jo                                                                                                                                                f
;-      3/4.10.1          Primary contairvnent Integrity /Drywelt Integrity                                                               Deleted                      No                 The latitude of this Special Test Exception is not required at                                      6'   i CPS.                                                                                                E i                                                                                                                                                                                                                                                                                             a y ~Ei n
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       -CLIOTON POWER STATION                                                                                                                            Page 14 of 16                                                           Sumary Disposition Matrix                                [bo m        \
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_ _ _ _ _                   __            - . _ _ _ _ _ _ _ - . _ - . . ..____- - . _ . _ .___                                                -~__.         .                --_.m...    . . _ . _ _ _ - _ . . _ _ _ _ . _ _ . _ . - . . _ .            _

==SUMMARY==
DISPOSITION MATRIX Retained / Criterion Old TS                                                    Title                           New Nunber           for inclusion         Bases for inclusion / Exclusion 3/4.10.2                                Rod Pattern Control System                         3.10.7                    Yes             See Note 4 3/6.10.3                                Shutdown Margin Demonstrations                     3.10.8                    Yes             See Note 4 S/6.10.4                                Recirculation Loops                                Deleted                   No              the Latitude of this Special Test Exception is not required at CPS.
3/4.10.5                                Training Startups                                  3.10.9                    Yes             See Note 4 3/4.10.6                                Special Instrtsnentation-Initial Core Loading      Deleted                   No              This Specification is only allowed during initlat core loading, which has been conpleted. Therefore, it is no longer applica-ble or needed and has been deleted.
M                                        DESIGN FEATURES                                    J 4                         Yes             See Note 5.
J G                                       ADMINISTRAf!VE CONTRot$                            M                         Yes             See Note 6.
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                                                                                                                                                                                                                                    *C CLINTON POWER STATION                                                                              Page 15 of 16                                                           Stenary Disposition MatrixM g 2 t2
                                                                                                                                                                                                                                    &R

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==SUMMARY==
DISPOSITION MATRIX NOTE 1:        DEFINITIONS This section provides definitions for several defined terms used throughout the remainder of Technical Specifications. They are provided to i mrove the meaning of certain terms. As such, direct application of the Technical Specification selection criteria is not appropeiste. However, only those definitions for defined terms that remain as a result of application of the selection criteria, will remain as definitions in this section of Technical Specifications.
NOTE 2:        SAFETY LIMITS /LSSS Application of Technical Specification selection criteria is not appropriate. However, Safety Limits and Limiting Safety System Settings (as part of Reactor Protection System Instrtsnentation) will be included in Technical Specifications as required by 10 CFR 50.36.
NOTE 3:       GENERIC 3.0/4.0 These Specifications provide generic guidance applicable to one or more Specifications. The Information is provided to facilitate understanding of Limiting Conditions for Operation and Surveillance Requirements. As such, direct application of the Technical Specification selection criteria is not appropriate.
However, the general requirements of 3.0/4.0 will be retained in Technical Specifications. Htsnan f actors ivrovements and certain technical imrovement' have been i glemented as agreed upon during NUMARC/NRC negotiations, NOTE 4:       SPECIAL TEST EXCEPTIONS These specifications are provided to allow relaxation of certain Limiting Conditions for Operation under certain specific conditions to allow testing and maintenance. They are directly related to one or more Limiting Conditions for Operation. Direct application of the Technical Specification selection criteria is not appropriate.      However, those special test exceptions, directly tied to Limiting Conditions for Operation that remain in Technical Specifications, will also remain as Technical Specifications. Those special test exceptions not applicable at CPS have been deleted.
COTE 5:       DESIGN FEATURES Application of Technical Specification selection criteria is not appropriate. However, Design Features will be included in Technical Specifications as required by 10 CFR 50.36.
NOTE 6:       ADMINISTRATIVE CONTROLS Arplication of Technical Specification selection criteria is not appropriate. However, Achinistrative Controls will be included in Technical Specifications as required by 10 CFR 50.36.
  !                                                                                                                                                                                2.
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                                                                                                                                                                               ';i e oe CLINTON POWER STATION                                                               Page 16 of 16                                                Sumary Disposition Matrix 0 st

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Page 30 of 65 ;
APPENDIX A 1
JUSTIFICATION FOR                              ;
i SPECIFICATION RELOCATION                          !
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I Enclosure 1 to U-60219t  ,
Page 31 of 6'  :
3/4.1.3.6 CONTROL ROD DRIVE HOUSING SUPPORT LCO Statement The control rod drive housing support shall be in place.
Discussion:
Control rod drive housing support supports control rod OPERABILITY by plant configuration management.            As such,    control rod OPERABILITY cannot be satisfied without the support being in place.
Without control rod OPERABILITY confirmed, appropriate Action Statements of the control rod OPERABILITY Specification must be entered. There is no need for duplicate requirements in a subsystem LCO. Relocation of this LCO is appropriate since plant configuration (the control rod housing support in place) would be controlled by post-maintenance procedures.
Comparison to Screenina Criteria:
: 1. The control rod drive housing support is neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
: 2. The control rod drive housing support is not capable of monitoring a process variable that is an initial condition of a DBA or transient analyses.
: 3. The control rod drive housing support is not part of a primary success path in the mitigatica of a DBA or transient. It does support the Control Rod OPERABILITY Specification which has been retained in the CPS Technical Specifications. As such, having the control drive housing support not in place impacts control rod OPEPABILITY, and appropriate actions are initiated which bound those actions that would be implemented because of the housing support being out of place. There is no need for duplicate actions.       Control rod drive housing support contribution to DBA and transient mitigation is preserved by the Control Rod OPERABILITY Specification          (LCO                                     3.1.3) requirements.
Probabilistic Risk Assessments        (PRAs)   address system risk contribution and identify systems which can be significant risk contributors to core damage and offsite releases.        Subcomponent risk contribution (in this case, the control rod drive housing support) is not addressed as the inoperability of the supported system (control rod drive system) and subsequent risk contribution is bounding.                                                                                                              .
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CLINTON POWER STATION        Page 1 of 35 Appendix A: Justification

l l                                                            Enclosure 1 to U-602196 Page 32 of 65 j

== Conclusion:==

Since the screening criteria have not been satisfied, the Control Rod Drive Housing Support LCO and Surveillances may be relocated to other   plant   controlled    documents  outside    the  Technical Specifications.
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r CLINTON POWER STATION       Page 2 of 25 Appendix A: Justification

Enclosure 1 to U-602196 Page 33 of 65 3/4.3.2      CONTAI! GENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM LCO Statement:
The containment       and reactor vessel      isolation control    system (CRVICS) channels shown in Table 3.3.2-1 shall be OPERABLE, with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.
Ambient Temperature Isolation Instrumentation and Differential Temperature Isolation Instrumentation.
Discussion:
The ambient temperature and differential temperature instruments proposed to be relocated are not assumed to function to mitigate any accident described in Chapters 6 or 15 of the Updated Safety Analysis Report.       These ambient and differential temperature instruments are provided only to detect and initiate isolation of a 25-gpm-equivalent steam leak.           However,   these instruments constitute only one method of determining steam leakage in their respective areas.      In addition to the temperature monitoring, excess reactor coolant leakage can be detected by low reactor water level, high process line flow, high differential flow, and various other plant specific methods.
4 Comparison to Screenina Criteria:
: 1. The Ambient Temperature and Differential Temperature Isolation Instruments are neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
: 2. The Ambient Temperature and Differential Temperature Isolation Instruments are neither used for, nor capable of, monitoring a process variable that is an initial condition of a DBA or transient analyses.
: 3. The Ambient Temperature and Differential Temperature Isolation Instruments are not used as parts of a primary success path in the mitigation of a DBA or transient. No pressure-temperature analyses,     radiation    dose   calculations,    or   equipment qualification parameters take credit for the operation of                        i these ambient or differential temperature instruments.               In addition, adequate redundancy is available to perform their functions by other methods.
Although the overall Isolation Instrumentation Function satisfies Criterion 3 of the NRC's Final Policy Statement on Technical Specification     Improvement,    these    ambient   temperature        and dif ferential temperature instruments are not assumed to function to mitigate any DBA or transient analyses.
CLINTON POWER STATION         Page 3 of 35 Appendix A: Justification

t Enclosure 1 to U-602190 Page 34 of 65

== Conclusion:==

Since the screening criteria have not been satisfied, the ambient and differential temperature instrument functions requirements of the    Improved   Technical    Specification   Isolation    Action Instrumentation may be relocated to other plant controlled documents outside the Technical Specifications.
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l CLINT^N POWER STATION      Page 4 of 35 Appendix A: Justification

l Enclosure 1 to U-602196 Page 35 of 65    .
3/4.3.3     EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION                                                  l LCO Statement:                                                                                                      !
J                                                                                                                          ;
)     The Emergency Core Cooling System (ECCS) actuation instrumentation 4
channels shown in Table 3.3.3-1 shall be OPERABLE, with their trip                                                  !
setpoints set consistent with the values shovn in the Trip Setpoint                                                 j
 !     column of Table 3.3.3-2 and with EMERGENCf CORE CX LING SYSTEM                                                      i I      RESPONSE TIME as shown in Table 3.3.3-3.                                                                            ;
3/4.3.3.A.2.h    ADS Trip System 1-Manual Inhibit ADS Switch 3/4.3.3.B.2.g    ADS Trip System 2-Manual Inhibit ADS Switch Discussion:
The ADS Manual Inhibit Switch allows the operator to defeat ADS                                                     i actuation as directed by the emergency operating procedures under conditions for which ADS would not be desirable.                       For . example,                                i during an ATWS event, low pressure ECCS activation would dilute                                                      ;
sodium pentaborate injected by the Standby Liquid Control (SLC)                                                     j System, thereby reducing the effectiveness of the SLC System                                                         ,
3 shutdown.                                                                                                            I 3
3 Comparison to Screenino Criteria:                                                                                   i i
)      1. The Manual Inhibit ADS Switch is neither used for, nor capable                                                 f
)           of,    detecting a significant abnormal degradation of the                                                     8 reactor coolant pressure boundary prior to a design basis accident (DBA).                                                                                                 j i
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: 2. The Manual Inhibit ADS Switch is neither used for, nor capable
.            of, monitoring a process variable that is an initial condition of a DBA or transient analyses.
: 3. The Manual Inhibit ADS Switch is not used as part of 'a primary success path in the mitigation of a DBA or transient. The inhibit feature was added to mitigate the consequences of an j           ATWS event which is not a DBA or transient. The switch does support the ADS system, which has been retained in the CES Technical Specifications.       The actions to be taken in the event that this switch is' positioned to defeat the ADS Jogic                                                  j bound those actions to be taken if the switch is inoperable.                                                   !
There is no need for duplicate actions.                   The ADS actuation                                     l instrumentation requirements (LCO 3.3.5.1) preserve ADS Manual                                                 l Inhibit Switch contribution to DBA and transient mitigation.                                                   l As discussed in Section 3.5 and summarized in Table 4-1 (item 112B) of NEDO-31466, the loss of the Manual Inhibit ADS Switch was found to be a non-significant risk contributor to core damage frequency and offsite releases. IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.
1 CLINTON POWER STATION          Page 5 of 35 Appendix A: Justification 1

Enclosure I to U-602196 Page 36 of 65

== Conclusion:==

  <   Since the screening criteria have not been satisfied, the portions                        '
l l    of the LCO and Surveillances applicable to the Manual Inhibit ADS l    Switch Function may be relocated to other plant controlled j     documents outside the Technical Specifications.
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,     CLINTON POWER STATION      Page 6 of 35 Appendix A: Justification

Enclosure I to U-602196 Page 37 of 65 3/4.3.6    CONTROL ROD BLOCK INSTRUMENTATION LCO Statement:                                                                       j l
The control rod block instrumentation channels shown in Table                        1 3.3.6-1 shall be OPERABLE, with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.
3/4.3.6.2   APRM Discussion:
The control rod block instrumentation is provided to prevent a control rod withdrawal error at power transient. APRMs utilize LPRM signals to provide information about the average core power.
As such,    they are not capable of providing the local power information necessary to mitigate a control rod withdrawal error transient. Therefore, these instruments are not used to mitigate a design basis accident (DBA) or transient.
Comparison to Screenino Criteria:
: 1. The APRM control rod block is neither used for, nor capable of,  detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
: 2. The APRM control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analyses.
: 3. The APRM control rod block signal is not a part of a primary success path in the mitigation of a DBA or transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 135) of NEDO-31466, the loss of the APRM control rod block function was                  l found to be a non-significant risk contributor to core damage                       l frequency and offsite releases. IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the control rod block LCO and Surveillances applicable to APRM instrumentation may be relocated to other plant controlled documents outside the Technical Specifications.
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i CLINTON POWER STATION        Page 7 of 35 Appendix A: Justification

Enclosure 1 to U-602196  )
Page 38 of 65   l I
3/4.3.6           CONTROL ROD BLOCK INSTRUMENTATION LCO Statement:
4 The control rod block instrumentation channels shown in Table                                  ,
3.3.6-1 shall be OPERABLE, with their trip setpoints set consistent                            !
with the values shown in the Trip setpoint column of Table 3.3.6-2.                            l 3/4.3.6.3          Source Range Monitors                                                       l Discussion:                                                                                    j The control rod block instrumentation is provided to. prevent a                                ,
control rod withdrawal error at power transient.               Source Range Monitor (SRM) signals are used to monitor neutron flux during                                 i refueling, shutdown, and startup conditions.               No design basis accident (DBA) or transient analysis takes credit for rod block signals initiated by the SRMs.
Comoarison to Screenina Criteria:                                                              !
]
  ,        1.       The SRM control rod block is neither used for, nor capable of,                        ;
l                   detecting a significant abnormal degradation of the reactor
;                   coolant pressure boundary prior to a DBA.
4          2.       The SRM control rod block instrumentation is not used to monitor a process variable that is an initial condition of a                          l DBA or transient analyses.
: 3.       The SRM control rod block signal is not a part of a primary                           ,
success path in the mitigation of a DBA or transient.                               li As discussed in section 3.5 and summarized in Table 4-1 (item 137)                             )
of NEDO-31466, the loss of the SRM control rod block function was
,          found to be a non-significant risk contributor to core damage                                  i
!          frequency and offsite releases.          IP has reviewed this evaluation,                      t i          considers it applicable to CPS, and concurs with the assessment.
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== Conclusion:==
I Since the screening criteria have not been satisfied, the control                              !
rod block LCO and Surveillances applicable to SRM instrumentation                              ;
i          may be relocated to other plant controlled documents outside the                               :
Technical Specifications.
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;                                                                                                           i CLINTON POWER STATION             Page 8 of 35 Appendix A: Justification

l Enclosure I to U-602196 Page 39 of 65    ;
3/4.3.6            CONTROL ROD BLOCK INSTRUMENTATION                                               ;
J LCO Statement:
The control rod block instrumentation channels shown in Table                                      !
3.3.6-1 shall be OPERABLE, with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.
3/4.3.6.4          Intermediate Range Monitors                                                    l h
Discussion:
The control rod block instrumentation is provided to prevent a control rod withdrawal error at power transient.                 Intermediate Range Monitors (IRMs) are provided to monitor the neutron flux levels during refueling, shutdown, and startup conditions.                    No design basis accident (DBA) or transient analysis takes credit for rod block-signals initiated by IRMs.                                                              ;
Comparison to Screenino Criteria:
: 1.       The IRM control rod block is neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
: 2.       The IRM control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analyses.                                                               ]
1
: 3.       The IRM control rod block signal is not a part of a primary success path in the mitigation of a DBA or transient.
l As discussed in section 3.5 and summarized in Table 4-1 (item 138) of NEDO-31466, the loss of the IRM control rod block function was found to be a non-significant risk contributor to core ' damage frequency and offsite releases.          IP has reviewed this evaluation, l           considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the control rod block LCO and Surveillances applicable to IRM instrumentation may be relocated to other plant controlled documents outside the Technical Specifications.
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CLINTON POWER STATION              Page 9 of 35 Appendix A: Justification

Endosure 1 to U-602196 Page 40 of 65 3/4.3.6     CONTROL ROD BLOCK INSTRUMENTATION LCO Statement:
The control rod block instrumentation channels shown in Table                          '
3.3.6-1 shall be OPERABLE, with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.
3/4.3.6.5   Scram Discharge Volume Discussion:
The control rod block instrumentation is provided to prevent a control rod withdrawal error at power transient. The purpose of measuring the scram discharge volume (SDV) water level is to ensure that there is sufficient volume remaining to contain the water discharged by the control rod drives during a scram, thus ensuring that the control rods will be able to insert fully. This rod block signal provides an indication to the operator that water is accumulating in the SDV and prevents further rod withdrawals. With                     >
continued water accumulation, a Reactor Protection System-initiated scram signal will occur. Thus, the SDV water level rod block signal provides an opportunity for the operator to take action to avoid a subsequent scram.       No design basis accident (DBA) or transient analysis takes credit for rod block signals initiated by the SDV instrumentation.                                                               l Comparison to Screenina Criteria:                                                      l
: 1. The SDV control rod block is neither used for, nor capable of, detecting a significant abnormal degradation of the reactor                      !
coolant pressure boundary prior to a DBA.                                        t
: 2. The SDV control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analyses.                                                       ,
: 3. The SDV control rod block signal is not a part of a primary success path in the mitigation of a DBA or transient.
As discussed in section 3.5 and summarized in Table 4-1 (item 139) of NEDO-31466, the loss of the SDV control rod block function was found to be a non-significant risk contributor to core d.amage frequency and offsite releases. IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the control rod block LCO and Surveillances applicable to SDV instrumentation may be relocated to other plant controlled documents outside the Technical Specifications.
CLINTON POWER STATION        Page 10 of 35 Appendix A: Justification

                             ..    --.               -. - .         -   -            ~.
l Enclosure I to UbO2196 4
Page 41 of 65 j      3/4.3.6    CONTROL ROD BLOCK INSTRUMENTATION LCO Statement:
The control rod block instrumentation channels shown in Table                           ;
3.3.6-1 shall be OPERABLE, with their trip setpoints set consistent                     ;
with the values shown in the Trip Setpoint column of Table 3.3.6-2.                     ,
3/4.3.6.6    Reactor Coolant System Recirculation Flow i
Discussion:                                                                             3 The control rod block instrumentation is provided to prevent a control rod withdrawal error at power transient.               Reactor recirculation flow provides input to the flow-biased setpoints of                       ,
the APRMs. No design basis accident (DBA) or transient analysis takes credit for rod block signals initiated by the APRMs.                              ,
l Comparison to Screenina Criteria:
: 1. The Reactor Coolant System (RCS) recirculation flow control                       !
rod block is neither used for, nor capable of, detecting a                        ,
  ,          significant abnormal degradation of the reactor coolant                           ;
I pressure boundary prior to a DBA.                                                 ,
: 2. The RCS recirculation flow control rod block instrumentation
;            is not used to monitor a process variable that is an initial                      i l            condition of a DBA or transient analyses.
a
: 3. The RCS recirculation flow control rod block signal is not a part of a primary success path in the mitigation of a DBA or transient.                                                                        t i
,      As discussed in section 3.5 and summarized in Table 4-1 (item 140) of NEDO-31466, the loss of the RCS recirculation    flow control rod block function was found to be a non-significant    risk contributor to core damage frequency and offsite releases.       IP has reviewed this evaluation, considers it applicable to CPS,    and concurs with                   ,
  ,     the assessment.

== Conclusion:==

i Since the screening critcria have not been satisfied, the control rod block LCO and Surveillances applicable to RCS recirculation i     flow instrumentation may be relocated to other plant controlled j     documents outside the Technical Specifications.
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A CLINTON POWER STATION         Page 11 of 35 Appendix A: Justification l

Enclosure I to U-602196 Page 42 of 65 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION LCO Statement:
The radiation monitoring instrumentation channels shown in Table 3.3.7.1-1 shall be OPERABLE, with their alarm / trip setpoints within the specified limits.
3/4.3.7.1.2     Area Monitors Discussion:                                                                            !
The area radiation monitors are used to indicate when the radiation in the new fuel storage vault, spent fuel storage pool, or main control room areas has exceeded its allowable setpoint.      There are no automatic functions that are performed by these instruments.
The instruments are not used to mitigate a design basis accident (DBA) or transient. Information provided by these instruments on                      I the radiation levels would have limited or no use in identify-ing/ assessing core damage.                                                           l l
Comparison to Screenina Criteria:
l
: 1. These monitors are neither used for, nor capable of, detecting                   l a significant abnormal degradation of the reactor coolant                       l pressure boundary prior to a DBA.
: 2. The monitored parameters are not assumed as initial conditions l
of a DBA or transient analysis that assumes the failure of, or presents a challenge to the integrity of a fission product barrier.
: 3. These monitors do not act as part of a primary success path in the mitigation of a DBA or transient that assumes the failure                   4 of, or presents a challenge to the integrity of a fission                       l product barrier.
As discussed in Section 3.5 and summarized in Table 4-1 (item 150)                    j of NEDO-31466, the loss of these monitors was found to be a non-significant risk contributor to core damage frequency and offsite releases. IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the Area Monitor LCOs and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
CLINTON POWER STATION         Page 12 of 35 Appendix A: Justification l

Enclosure 1 to U-602196   i Page 43 of 65 ;
3/4.3.7.2 SEISMIC MONITORING INSTRUMENTATION LCO Statement:                                                                              '
The seismic monitoring instrumentation shown in Table 3.3.7.2-1 shall be OPERABLE.                                                                          .
Discussion:
In the event of an earthquake, seismic instrumentation is required to permit comparison of the measured response to that used in the                          ;
design basis of the facility to determine if plant shutdown is                              :
required pursuant to Appendix A of 10 CFR Part 100.      There is no                       f automatic action that these instruments perform during a seismic event. Since this is determined after the event has occurred, it                        ,
has no bearing on the mitigation of any design basis accident                              ,
(DBA). The magnitude of the earthquake can also be obtained from the National Earthquake Information Service or other sources.
Comparison to Screenino Criteria:                                                          ;
: 1. These instruments are neither - used for, nor capable of,                            l detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.                                            I
: 2. These instruments do not monitor a process variable that is an                       !
initial condition to a DBA or transient analysis.                                    l
: 3. These instruments do not act as part of a primary success path in the mitigation of a DBA or transient.
I As discussed in Section 3.5 and summarized in Table 4-1 (item 151) l    of NEDO-31466, the loss of seismic monitoring instrumentation was l     found to be a non-significant risk contributor to core damage l     frequency and offsite releases. IP has reviewed this evaluation, I    considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the Seismic Monitoring LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
l CLINTON POWER STATION        Page 13 of 35 Appendix A: Justification l                ._

Enclosure 1 to U-602196   ;
Page 44 of 65 ;
;   3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION LCO Statement:                                                                              !
The meteorological monitoring instrumentation channels shown in Table 3.3.7.3-1 shall be OPERABLE.
Discussion:                                                                                 {
Meteorological instrumentation is used to measure environmental parameters which may affect distribution of fission products and gases following a design basis accident (DBA), but it is not an input assumption for any DBA analysis and does not mitigate the accident.      There is no automatic action that these instruments perform during any event. Meteorological information is required to evaluate the need for initiating protective measures to protect the health and safety of the public in the event of an accident.                            .
However, this information can be obtained from the National Weather Service or other sources.                                                                   j Comparison to Screenina Criteria:
: 1. These instruments are neither used for,          nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.                                            j
: 2. These monitored      parameters--wind    direction,    speed,    air                 '
temperature, and air temperature differences-are not process variables that are initial conditions in a DBA or transient                          I analysis.
: 3. These instruments do not act as a part of a primary success                          [
path in the mitigation of a DBA or transient.                                        !
As discussed in Section 3.5 and summarized in Table 4-1 (item 152)                           i of      NEDO-31466,     the   loss    of    meteorological      monitoring                  l instrumentation was found to be a non-significant risk contributor to core damage frequency and offsite releases.           IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

,    Since    the screening criteria have not been satisfied,                the Meteorological Monitoring LCO and Surveillance may be relocated to other     plant    controlled   documents     outside    the    Technical Specifications.
1
~
CLINTON POWER STATION         Page 14 of 35 Appendix A: Justification

Enclosure 1 to U-602196 Page 45 of 65 7
3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION                                            !
l l   LCO Statement:
The accident monitoring instrumentation channels shown in Table                          ,
3.3.7.5-1 shall be OPERABLE.                                                             :
Discussion:
Each individual accident monitoring parameter has a specific                             f purpose, however, the general purpose for all accident monitoring                        l instrumentation is to provide sufficient information to assess                           i
-     plant response in the event of an accident; i.e., automatic safety
;     systems are performing properly and deviations from expected                             '
accident course are minimal.
4                                                                                              ;
comparison to screenino Criteria:
The NRC position on application of the      screening criteria to post-accident monitoring instrumentation is      documented in letter dated 1     May 7, 1988, from T.E. Murley (NRC) to      R.F. Janecek (BWROG). The position      taken    was   that       the post-accident   monitoring instrumentation table list should contain, on a plant specific                           +
i     basis, all Regulatory Guide 1.97 Type A instruments specified in                         E the plant's SER on Regulatory Guide 1.97, and all Regulatory Guide                       .
1.97 Category 1 instruments. Accordingly, this position has been applied to the CPS Regulatory Guide 1.97 instruments.              Those                 ;
instruments meeting this criteria have remained in Technical                             !
Specifications, and those instruments not meeting the criteria have                      !
been     relocated from    the  Technical    Specifications   to   plant                 !
controlled documents.
The following summarizes the application of the NRC position to CPS.
* From SER Supplement 5, dated January,1986, 

==Subject:==
SER Related to the Operation of Clinton Power Station Unit 1 and Regulatory Guide 1.97:
)           Tvoe A Variables l
: 1. Reactor Vessel Pressure
: 2. Reactor Vessel Water Level
: 3. Suppression Pool Water Level
: 4. Suppression Pool Water Temperature
: 5. Drywell Pressure 4            6. Drywell/ Containment Hydrogen and Oxygen Concentration Analyzer and Monitor                                                         i I
j 1                                                                                               )
i     CLINTON POWER STATION         Page 15 of 35 Appendix A: Justification 1

i Enclosure 1 to U-602196 i Page 46 of 65 Other Tvoe. Cateoorv 1 Variables
: 1. Containment Pressure
: 2. Containment /Drywell High Range Gross Gamma Radiation                   ,
Monitors
: 3. Primary Containment Isolation Valve Position Indication Conclusion Since    the   screening  criteria  have   not  been  satisfied     for instruments which do not meet Regulatory Guide 1.97 Type A variable requirements or Category 1 variable requirements, their associated                   ,
LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications. The instruments to be relocated are as follows:
: 1. Drywell Air Temperature
: 2. Containment Temperature                                                        ,
: 3. Safety / Relief Valve Acoustic Monitor
: 4. HVAC Stack High Range Radioactivity Monitor
: 5. SGTS Exhaust High Range Radioactivity Monitor 4
i l
1 CLINTON POWER STATION        Page 16 of 35 Appendix A: Justification

Enclosure 1 to U-602196 -
Page 47 of 65 ,
3/4.3.7.7 TRAVERSING IN-CORE PROBE i
LCO Statement:
l
.       The traversing in-core probe system shall be OPERABLE, with:
Discussion:                                                                           l The traversing in-core probe (TIP) system is used only for
 ;      calibration of the LPRM detectors. The TIP System is positioned axially and radially throughout the core to calibrate the local power range monitors (LPRMs) . When not in use, the TIP instruments are retracted into a storage position inside the drywell wall
:       penetrations. The TIP System supports the OPERABILITY of the LPRMs. With LPRM OPERABILITY addressed, there is no need to address the TIP System in the Technical Specifications.
Comparison to screenino Criteria:
: 1. The TIP System is neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
: 2. The TIP System is not used to monitor a process variable that                   ;
is an initial condition of a DBA or transient analyses.
: 3. The TIP System is not a part of a primary success path in the mitigation of a DBA or transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 183)                    .
of NEDO-31466, the loss of the TIP System was found to be a non-                      !
significant risk contributor to core damage frequency and offsite                     ;
releases. IP has reviewed this evaluation, considers it applicable                    ;
to CPS, and concurs with the assessment.
B

== Conclusion:==

Since the screening criteria have not been satisfied, the TIP System LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
1 CLINTON POWER STATION        Page 17 of 35 Appendix A: Justification

Enclosure 1 to U-602196 Page 48 cf 65 r
3/4.3.7.8 CHLORINE DETECTION SYSTEM LCO Statement:
a   Two independent chlorine detection channels shall be OPERABLE, with their     trip  setpoints adjusted    to  actuate   at  a   chlorine concentration of 5 5 ppm.
Discussion:
The Chlorine Detection System is used to isolate the control room upon detection of a high concentration of chlorine. The chlorine release would not be a result of a design basis accident (DBA) or transient; thus, the instruments do not perform any required function during a design basis event. Amendment No. 12 to the CPS Technical Specifications incorporated provisions that this LCO would no longer be applicable after all chlorine containers having                        ,
4 a capacity of 100 pounds or greater are removed from the site.
Because this condition has been met, this LCO is no longer required to be met at CPS.
Comparison to Screenino Criteria:
[
: 1. The Chlorine Detection System is neither used for, nor capable                      >
of,   detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.                                   l
: 2. The Chlorine Detection System is not used to monitor a process variable that is an initial condition of a DBA or transient                         ,
analyses.
: 3. The Chlorine Detection System is not part of a primary success                      I path in the mitigation of a DBA or transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 184) of NEDO-31466, the loss of the Chlorine Detection System was found                         ;
to be a non-significant risk contributor to core damage frequency                          I and offsite releases. IP has reviewed this evaluation, considers                        I it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satir.fied, the Chlorine Detection System LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
CLINTON POWER STATION       Page 18 of 35 Appendix A: Justification 1

l Enclosure I to U-602196 Page 49 of 65 3/4.3.7.10      LOOSE-PART DETECTION SYSTEM LCO Statement:
The loose-part detection system shall be OPERABLE.
Discussion:
The Loose-Part Detection System is used to detect loose parts in the reactor vessel. The instrumentation does not indicate that there is a degradation in the primary pressure boundary but indicates that there might be a remote chance of damage to a component due to a loose part. The potential of fuel failure due to fuel bundle flow blockage from a lost part will be detected by the radiation monitors in the offgas stream.                                          ;
Comparison to Screenino Criteria:
: 1. The Loose-Part Detection System is neither used for, nor capable of, detecting a significant abnormal degradation of                      ,
the reactor coolant pressure boundary prior to a design basis accident (DBA).                                                                  r
: 2. The Loose-Part Detection System is not used to monitor a process variable that is an initial condition of a DBA or                        i transient analyses.
: 3. The Loose-Part Detection System is not part of a primary success path in the mitigation of a DBA or transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 187)                    ,
of NEDO-31466, the loss of the Loose-Part Detection System was found to be a non-significant risk contributor to core damage frequency and offsite releases. IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.
I

== Conclusion:==

Since the screening criteria have not been satisfied, the Loose-                      ,
Part Detection System LCO and Surveillances may be relocated to                       ,
other    plant   controlled    documents   outside   the   Technical Specifications.
i 1
CLINTON POWER STATION        Page 19 of 35 Appendix A: Justification l

Enclosure I to U-602196 Page 50 of 65 '
3/4.3.7.11       MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING INSTRUMENTATION                                       :
i                                                 LCO Statement:
At least one main condenser offgas treatment system explosive gas monitoring instrumentation channel shall be OPERABLE, with its alarm / trip setpoint set to ensure that the limits of Specification 3.7.8.1 are not exceeded.
Discussion:                                                                           ,
The explosive gas monitor Specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the main condenser offgas treatment system is adequately                          '
monitored, which will help ensure that the concentration is                           '
maintained below the flammability limit of hydrogen. However, the offgas system is designed to contain detonations and will not affect the    function of any safety related equipment.            The concentration of hydrogen in the offgas stream is not an initial assumption of any design basis accident (DBA) or transient analysis.
Comparison to Screenino Criteria:
: 1. The explosive gas mixture indication is neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
: 2. The explosive gas mixture is not a process variable that is an initial condition of a DBA or transient analyses.
: 3. The explosive gas mixture indication is not utilized in any
 ,                                                         capacity in a primary success path in the mitigation of a DBA or transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 306) of NEDO-31466, an explosive gas mixture in the Offgas Treatment System was found to be a non-significant risk contributor to core damage frequency and offsite releases.        IP has reviewed this evaluation, considers it applicable to CPS, and concurs with this assessment.
1

== Conclusion:==

Since the screening criteria have not been satisfied, the Main Condenser Offgas Treatment System Explosive Gas Monitoring Instrumentation LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.                        l
^
l i
CLINTON POWER STATION       Page 20 of 35 Appendix A: Justification

l Enclosure I to U-602196 Page 51 of 65 3/4.3.9.2 FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM LCO Statement:
The Feedwater System / Main Turbine Trip System shall be OPERABLE.
Discussion:
The Feedwater System / Main Turbine Trip on Reactor Vessel Water Level-411gh, Level 8 is used in the Design Basis transient analysis                    '
for plants that do not have a scram from Reactor Protection System Reactor Vessel Level 8. Clinton (like all other BWR-6 plants) has                      l a direct scram on Reactor Vessel Water Level 8. The Design Basis                       '
transient analysis does not require the scram that would be received from the trip of the main turbine. Consequently, this LCO                   i does not serve any primary safety function (i.e., detection or mitigation of a design basis accident (DBA) or transient).
Comparison to Screenino Criteria:
: 1. The Feedwater System / Main Turbine Trip System is neither used for,   nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
: 2. The Feedwater System / Main Turbine Trip System is not capable of monitoring a process variable that is an initial condition of a DBA or transient analyses.
: 3. The Feedwater System / Main Turbine Trip System is not part of a   primary success path    in the mitigation of a DBA or transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 194) of NEDO-31466, the loss of the Feedwater System / Main Turbine Trip System was found to be a non-significant risk contributor to core damage frequency and offsite releases.        IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the Feedwater System / Main Turbine Trip System LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
CLINTON POWER STATION       Page 21 of 35 Appendix A: Justification

Enclost e I to U-60219 Page 52 of 6 3/4.3.9.3 SUPPRESSION POOL MAKEUP SYSTEM (SPMS)
LCO Statement:
The Suppression Pool Makeup System (SPMS) Mode Switch Permissive shall be OPERABLE.
Discussion:
The SPMS Mode Switch Permissive Function is an operational function l     only and is not considered in any design basis accident (DBA) or I
transient analysis. In addition, this switch permissive function is   controlled under administrative controls to assure the appropriate position of the switch is maintained.
Comparison to Screenino Criteria:
: 1. The SPMS Mode Switch Permissive is neither used for, nor capable of, detecting a significant abnormal degradation of j           the reactor coolant pressure boundary prior to a DBA.
: 2. The SPMS Mode Switch Permissive is not used to monitor a process variable that is an initial condition of a DBA or transient analyses.
: 3. The SPMS Mode Switch Permissive is not part of a primary success path in the mitigation of a DBA or transient.

== Conclusion:==

Since the screening criteria have not been satisfied, the SPMS Mode                     ,
Switch Permissive Function may be relocated to other plant                              i controlled documents outside the Technical Specifications.
i l
l l
CLINTON POWER STATION        Page 22 of 35 Appendix A: Justification

4 $
Endosure 1 to U-602196 a
Page 53 of 65 3/4.3.10    NUCLEAR SYSTEM PROTECTION SYSTEM-SELF TEST SYSTEM LCO Statement:
The Self Test SYSTEM (STS) of the Nuclear System Protection System shall be OPERABLE and operating in the fully automatic mode.
Discussion:
The primary purpose of the Self Test System is to enhance the availability of the Nuclear System Protection System by optimizing the time to detect and determine the location of a failure in the functional system. The Self Test System is used for post-maintenance testing and to augment conventional testing methods to perform various surveillance testing functions.
Comparison to Screenina criteria:
: 1. The Nuclear System Protection System-Self Test System is                                       i neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
: 2. The Nuclear System Protection System-Self Test System is not used to monitor a process variable that is an initial condition of a DBA or transient analyses.
: 3. The Nuclear System Protection System-Self Test System is not part of a primary success path in the mitigation of a DBA or
;           transient.
As discussed in Section 6 and summarized in Table 4-1 (item 312) of NEDO-31466, Supplement 1, the loss of the Nuclear System Protection System-Self Test System was found to be a non-significant risk contributor to core damage frequency and offsite releases.                  IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the Nuclear System Protection System-Self Test System LCO and Surveillances may be relocated to other plant controlled documents outside the                                         ,
Technical Specifications.                                                                             I
 !                                                                                                          l i
3 l
CLINTON POWER STATION      Page 23 of 35 Appendix A: Justification

i i
Enclosure I to U-6021%
Page 54 of 65 I 3/4.4.4      CHEMISTRY                                                                       !
l LCO Statement:
1 The chemistry of the reactor coolant system shall be maintained                              i within the limits specified in Table 3.4.4-1.                                                l t
Discussion:                                                                                  l Poor coolant water chemistry contributes to the long-term                                    ,
degradation of system materials of construction and thus is not of                           l immediate importance to the plant operator. Reactor coolant water                            i chemistry is monitored for a variety of reasons. One reason is to                            !
reduce the possibility of failures in the Reactor Coolant System                             -
pressure boundary caused by corrosion. Severe chemistry transients have resulted in failure of thin walled LPRM instrument dry tubes                            ,
in a relatively short period of time. However,these LPRM dry tube                            !
failures result in loss of the LPRM function and are readily                                 i i       detectable. In sunmary, the chemistry monitoring activity is of a l       long-term preventative purpose rather than mitigative.                                       l 1
l Comparison to Screenino Criteria:                                                            '
l
: 1. Reactor coolant water chemistry is neither used for, nor                                :
capable of, detecting a significant abnormal degradation of                             ;
the reactor coolant pressure boundary prior to a design basis                           i accident (DBA).                                                                         !
: 2. Reactor coolant water chemistry is not used to monitor a                                ;
process variable that is an initial condition of a DBA or                               i transient analyses.                                                                     !
: 3. Reactor coolant water chemistry is not supportive of any                                 ;
primary success path in the mitigation of a DBA or transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 211)                             l of NEDO-31466, the reactor coolant water chemistry was found to be a non-significant risk contributor to core damage frequency and offsite releases.      IP has reviewed this evaluation, considers it                           ,
applicable to CPS, and concurs with the assessment.                                            !

== Conclusion:==

Since the screening criteria have not been satisfied, the Reactor Coolant System Chemistry LCO and Surveillances may be relocated to other   plant      controlled   documents   outside    the       Technical Specifications.
i i
l i
i CLINTON POWER STATION         Page 24 of 35 Appendix A: Justification

s
  . .                                                                                      i Enclosure 1 to U-60219 3 Page 55 cf 6 3/4.4.8      STRUCTURAL INTEGRITY LCO Statement:
The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.8.                           ;
Discussion:
The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained throughout the component's life.          Other Technical Specifications require important systems to be OPERABLE (for example, ECCS 3/4.5.1) and in a ready state for mitigative action.                    .
This Technical Specification is more directed toward prevention of component degradation and continued long-term maintenance of acceptable structural conditions.      Hence, it is not necessary to retain this Specification to ensure immediate OPERABILITY of safety systems.
Comparison to Screenino Criteria:
: 1. The inspections stipulated by this Specification are neither used for, nor capable of, detecting a significant abnormal l          degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
: 2. The inspections stipulated by this Specification do not                         !
monitor process variables that are initial assumptions in a DBA or transient analysis.
: 3. The ASME Code Class 1, 2, and 3 components inspected per this Specification are assumed to function to mitigate a DBA.
Their capability to perform this function is addressed by                        ,
other Technical Specifications. This Technical Specification,                   i however, only specifies inspection requirements for these                        ;
components. Therefore, Criterion 3 is not satisfied.
As discussed in Section 3.5 and summarized in Table 4-1 (item 216) of NEDO-31466, the assurance of OPERABILITY of the entire system as verified in the system OPERABILITY Specification dominates the risk contribution of the system. The lack of a long-term assurance of structural integrity as stipulated by this Specification was found to be a non-significant risk contributor to core damage frequency and offsite releases. IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.                                 j l                                                                                           \

== Conclusion:==

1 Since    the   screening  criteria   have  not  been  satisfied,    the Structural Integrity LCO and Surveillances may be relocated to other     plant    controlled   documents   outside   the   Technical Specifications.
CLINTON POWER STATION         Page 25 of 35 Appendix A: Justification

Enclosure 1 to U402196 Page 56 of 65 3/4.7.4     SNUBBERS LCO Statement:
All snubbers shall be OPERABLE.
Discussion:
Snubbers are included in the plant design to ensure that the structural integrity of the reactor coolant system and other safety related systems are maintained during and after a seismic or other dynamic loading event. The snubbers are considered a part of the piping system. They serve as an aid to prevent piping failure, but do not mitigate piping failure should it occur. Also, the failure of a snubber on a particular pipe cannot, by itself, cause the pipe to fail. Consequently, the snubbers do not meet any of the criteria, since they are not utilized as part of the primary                          +
success path in detecting or mitigating the consequences of a design basis accident (DBA) or transient event. Additionally, the surveillance and maintenance of the snubbers can be controlled by sources other than the plant Technical Specifications.
Comparison to Screenina criteria:                                                     l
: 1. Snubbers are neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant                          -
pressure boundary prior to a DBA.
i
: 2. Snubbers are not used to monitor a process variable that is an initial condition of a DBA or transient analyses.
: 3. Snubbers are not part of a primary success path in the mitigation of a DBA or transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 266) of NEDO-31466, the loss of snubbers was found to be a non-significant risk contributor to core damage frequency and offsite releases. IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the Snubber LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
l l
l l
CLINTON POWER STATION       Page 26 of 35 Appendix A: Justification

a Enclosure I to U402196 Page 57 of 65 3/4.7.5       SEALED SOURCE CONTAMINATION i              LCO Statement:
Each sealed source containing radioactive material either in excess                          j of 100 microcuries of beta and/or gamma emitting material or 10                             i microcuries of alpha emitting material shall be free of greater                             '
i 3
than or equal to 0.005 microcuries of removable contamination.
1 j              Discussion:                                                                                  l 1                                                                                                          l The limitations on sealed source contamination are intended to                              )
1             ensure that the total body or individual organ irradiation doses do                          !
d not exceed allowable limits in the event of ingestion or                                    ;
inhalation. This is done by imposing a maximum limitation of less                          i than or equal to 0.005 microcuries of removable contamination on                            i each sealed source.           This requirement and the associated                           l surveillance requirements bear no relation to the conditions or limitations which are necessary to ensure safe reactor operation.                         ;
Comparison to Screenina Criteria:
: 1. Sealed source contamination is neither used for, nor capable                         i of,   detecting a significant abnormal degradation of the                            f 1                     reactor coolant pressure boundary prior to a design basis                            l accident (DBA).                                                                      l 2
: 2. Sealed source contamination is not a process variable that is an initial condition of a DBA or transient analyses.
(
j
: 3. Sealed source contamination is not used in any part of a                             l primary success path in the mitigation of a DBA or transient.                        j i
As discussed in Section 3.5 and. summarized in Table 4-1 (item 267) of NEDO-314 66, the sealed source contamination being not within                           )
;               limits was found to be a non-significant risk contributor to core damage frequency and offsite releases.           IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.
m

== Conclusion:==

1
)               Since the screening criteria have not been satisfied, the Sealed Source Contamination LCO and Surveillances may be relocated to 4               other     plant    controlled    documents   outside      the  Technical Specifications.
1 i
I 1              CLINTON POWER STATION           Page 27 of 35 Appendix A: Justification

Enclosure 1 to U-60219<
Page 58 of 6:
3/4.8.4.1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES                                                                   ;
LCO Statement:                                                                           j Primary and backup containment penetration conductor overcurrent protective devices associated with each primary containment                              i electrical penetration circuit shall be OPERABLE.       The scope of                    i these protective devices excludes those circuits for which credible                     e fault currents would not exceed the electrical penetrations' design ratings.                                                                                1 Discussion:
The primary feature of these protective devices is to open the l                  control and/or power circuit whenever the load conditions exceed the preset current demands.        This is to protect the circuit conductors against damage or failure due to overcurrent heating effects.
The continuous monitoring of the operating status of the overcurrent protective devices is impracticable and not covered as part of the control room monitoring, except after trip condition                        ,
indication.
In the event of failure of this protective device to trip the i
circuit, the upstream protective device is expected to operate and l                   isolate the faulty circuit.       Thus, the upper level    (back-up) protection will protect the circuit conductors. In the worst-case fault condition, a single division of protective functions can be                      i lost. However, this scenario is covered under a single failure                     ;
criterion.
The overcurrent protective devices ensure the pressure integrity of the containment penetration. With failure of the device it is postulated that the wire insulation will degrade, resulting in a containment leak path during a LOCA.       However, the protection provided by these devices is not a process variable and is not                           l considered as part of the primary success path.          Containment                    l penetration degradation will be identified during the normal I                   containment leak rate tests required by 10 CFR Part 50, Appendix J.
Comparison to Screenino Criteria:
: 1. The containment penetration conductor overcurrent protective                      l devices are neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
: 2. The containment penetration conductor overcurrent protective devices do not monitor a process variable that is an initial condition of a DBA or transient analyses.
: 3. The specific circuits of the containment penetration conductor overcurrent protective devices are not part of a primary success path in the mitigation of a DBA or transient.
CLINTON POWER STATION        Page 28 of 35 Appendix A: Justification                     ;
I l

Enclosure 1 to U-6021% !'
Page 59 of 65 i As discussed in Section 3.5 and summarized in Table 4-1 (item 276)                             ;
of NEDO-31466,     loss of the containment penetration conductor                                ;
overcurrent protective devices was found to be a non-significant-                              l risk contributor to core damage frequency and offsite releases. IP                             i has reviewed this evaluation, considers it applicable to CPS, and                              ;
concurs with the assessment.                                                                   ,

== Conclusion:==

Since  the     screening   criteria  have      not been  satisfied,    the                     !
l          Containment Penetration Conductor Overcurrent Protective Devices                               (
l LCO and Surveillances may be relocated to other plant controlled                               j documents outside the Technical Specifications.                                                i i
l 4
l i
l                                                                                                         i 1
l 1
CLINTON POWER STATION         Page 29 of 35 Appendix A: Justification
                                                                                                  . ~ . .

Enclosure I to U-602196 Pcge 60 of 65 3/4.8.4.2 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION LCO Statement:
The thermal overload protection of each valve in safety systems with a bypass device (s) integral with the motor starter shall be bypassed continuously for those directions for which the valve performs an active safety function.
Discussion:
For valves with thermal overload protection (i.e. , trip on overload condition), the valve function should be accomplished prior to overload trip. The overload protection for these valves is meant to take precedence over the valve function.            If an overload condition occurs during valve operation, the electrical circuit will open to protect the equipment. In case of failure of overload                   l protection operation to disconnect the load, the equipment may suffer potential damage. This may impact the OPERABILITY of the system containing the valve.       Accordingly, the system LCO would address the overall system OPERABILITY, and not the OPERABILITY of a support system. Additionally, the surveillance and maintenance of the motor operated valves thermal overlo d protection can be controlled     by    sources   other   than    the   plant    Technical Specifications.
Comparison to Screening Criteria:
: 1. Motor operated valve thermal overload protection is neither used for, nor capable of, detecting a significant abnormal
        . degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
: 2. Motor operated valve thermal overload protection does not monitor a process variable that is an initial condition of a DBA or transient analyses,                                                        j
: 3. Actuation  of   a motor   operated valve's thermal overload                       j protection is not part of a primary success path in the mitigation of a DBA or transient. The supported system (e.g.,
ECCS) may be part of a success path and is then retained in the Technical Specifications. However, motor operated valve thermal overload protection retention in the Technical Specifications is not necessary as its function is confirmed in the OPERABILITY of the supported system.
Probabilistic Risk Assessments         (PRAs)   address   system     risk contribution and identify systems which can be significant risk contributors to core damage and offsite releases.          Subcomponent                  '
l risk contribution from motor operated valve thermal overload protection malfunction is not addressed as the inoperability of the                      ;
supported system and subsequent risk contribution is bounding.                           !
l CLINTON POWER STATION        Page 30 of 35 Appendix A: Justification

- e Enclosure I to U-60219(
Page 61 of 6.

== Conclusion:==

Since the screening criteria have not been satisfied, the Motor Operated Valve Thermal Overload Protection LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications, i
l 4
i CLINTON POWER STATION    Page 31 of 35 Appendix A: Justification

Enclosure I to U-602196. ' i 3/4.9.5     COMMUNICATIONS                                             Page 62 0f 65     :
i LCO Statement:                                                                           j Direct communication shall be maintained between the control room and refueling platform personnel.                                                        :
Discussion:
Communication between the control room and refueling- floor                              !
personnel is maintained to ensure that refueling personnel can be                        I promptly informed of significant changes in the plant status or core reactivity condition during refueling.      The communications                      i allow for coordination of activities that require interaction                           i between the control room and refueling floor personnel (such as the insertion of a control rod prior to loading fuel). However, the                         j refueling system design accident or transient response does-not                          i take credit for communications and is designed to ensure safe refueling operations.
Comparison to Screenino Criteria:                                                       l l
: 1. Communications during any mode of plant operation is neither used for, nor capable of, detecting a significant abnormal                        j degradation of the reactor coolant pressure boundary prior to                     '
a design basis accident (DBA).
: 2. Communications during any mode of plant operation is not used                     ,
to indicate status of, or monitor a process variable that is                      j an initial condition of a DBA or transient analyses.                              !
: 3. Communication during refueling operations does not contribute                     l to a primary success path in the mitigation of a DBA or                           l transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 286) of NEDO-31466, the loss of communication was found to be a'.non-                        ,
significant risk contributor to core damage frequency and offsite                       l releases. IP has reviewed this evaluation, considers it applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the Communications LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
l l
1 CLINTON POWER STATION      Page 32 of 35 Appendix A: Justification-                    l

        . O Enclosure 1 to U-602196 3/4.9.6         FUEL HANDLING EQUIPMENT                                      Page 63 of 65   ;
LCO Statement:
3.9.6.1         The refueling platform shall be OPERABLE and used for                         ;
handling fuel assemblies or control rods within the                           ;
reactor pressure vessel.
3.9.6.2        The auxiliary platform shall be OPERABLE.                                     !
2 Discussion:
OPERABILITY of the refueling equipment (refueling and auxiliary platforms) ensures that only the proper hoists of the refueling and                           ,
auxiliary platforms will be used to handle fuel within the reactor                           j pressure vessel or fuel pool, hoists have sufficient load capacity                           i
 +
for handling fuel assemblies and/or control rods and the core                                i internals and pressure vessel are protected from excessive lifting                          j force if they are inadvertently engaged during lifting operations.                          ;
4          Although the interlocks designed to provide the above capabilities                           .
can prevent damage to the fuel handling equipment and core                                  !
internals, they are not assumed to function to mitigate the consequences of a design basis accident (DBA).                                              :
Comparison to Screenino Criteria:                                                           i
: 1. The fuel handling equipment and associated instrunentation are                          ,
                                                                                                         ~
neither used for, nor capable of, detecting a significant                               ,
abnormal degradation of the reactor coolant pressure boundary                           !
prior to a DBA.                                                                         {
: 2. The fuel handling equipment and associated instrumentation are                           l not used to monitor a process variable that is an initial                               ;
.                condition of a DBA or transient analyses.                                               i 9
: 3. The fuel handling equipment and associated instrumentation are not part of a primary success path in the mitigation of a DBA or transient.
;           As discussed in Section 3.5 and summarized in Table 4-1 (item 287)
'           of    NEDO-31466,       the    refueling     equipment and    associated instrumentation was found to be a non-significant      risk contributor to core damage frequency and offsite releases.          IP has reviewed this evaluation, considers it applicable to CPS,       and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the Fuel Handling Equipment LCOs and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
CLINTON POWER STATION           Page 33 of 35 Appendix A: Justification

Enclosure 1 to U-602196 Pege 64 of 65 3/4.9.7      CRANE TRAVEIr-SPENT FUEL STORAGE POOL, UPPER CONTAINMENT FUEL POOL, AND NEW FUEL STORAGE VAULT LCO Statement _;_
Loads in excess of 1000 pounds shall be prohibited from travel over fuel assemblies in the spent fuel storage pool racks, upper containment fuel pool racks or new fuel storage vault racks.
Discussion:
The restriction on movement of loads in excess of the nominal weight of a fuel assembly over other fuel assemblies in the storage pools ensures that in the event the load is dropped, the activity release will be limited to that contained in a single fuel assembly and any possible distortion of the fuel in the storage racks will not result in a critical array. Administrative monitoring of loads moving over the fuel storage racks serves as a backup to the crane interlocks.
Altbaugh this LCO supports the maximum refueling accident assumption in the DBA, these types of limitations are adequately controlled by administrative controls. Therefore, the criteria for Technical Specification retention are not satisfied.
Comparison to Screenino Criteria:
: 1. The crane travel load limits are neither used for, nor capable of,   detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).                                                                    ,
i
: 2. The maximum severity assumed for the fuel handling DBA is                          l limited by the load limits placed on the crane travel. These crane travel limits are not, however, process variables monitored and controlled by the operator.           They may be interlocks and/or physical stops and are           addressed by administrative controls. Criterion 2 is thus not satisfied.
l
: 3. The crane travel load limits are not a structure, system, or                       i component that is part of the primary success path and which functions or actuates to mitigate a DBA.
Traditional Probabilistic Risk Assessments (PRAs) do not review risks associated with the spent fuel storage pool.

== Conclusion:==

Since the screening criteria have not been satisfied, the Crane Travel-Spent Fuel Storage Pool, Upper Containment Fuel Pool, and New Fuel Storage Vault LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifica-tions.
1 CLINTON POWER STATION        Page 34 of 35 Appendix A: Justification

f Enclosure 1 to U-6021     '
3/4.9.12     INCLINED FUEL TRANSFER SYSTEM                                                    Page 65 of 65 LCO Statement:                                                                                                  ,
l The inclined fuel transfer system (IFTS) may be in operation                                                    {
provided that.
I
: a. The access doors of all rooms through which the transfer system penetrates are closed and locked.
: b. All access door interlocks are OPERABLE.
: c. The blocking valve located in the fuel building IFTS hydraulic power unit is OPERABLE.                                                                                    ;
: d. At least one IFTS carriage position indicator is OPERABLE at each carriage position and at least one liquid level sensor is                                            ;
OPERABLE.                                                                                                  :
e
: e. Any keylock switch that provides IFTS access control-transfer                                             !
system lockout is OPERABLE.
Discussion:
i The IFTS transfers fuel from the secondary containment                                  (fuel building) into primary containment (upper fuel pool).- The purpose                                            i of  the IFTS      Specification is to limit personnel access to                                                ;
potentially high-radiation areas of the system.                       This requirement                         !
is not an assumption of any design basis accident (DBA), but helps                                             '
to ensure that 10 CFR 20 limits are not exceeded.                                                             l Comoarison to Screenino Criteria:
: 1. The IFTS is neither used for, nor capable of, detecting a                                                  !
significant abnormal degradation of the reactor coolant                                                    1 pressure boundary prior to a DBA.                                                                          ;
1
: 2. The IFTS is not used to monitor a process variable that is an initial condition of a DBA or transient analyses.
: 3. The IFTS is not part of a primary success path in the mitigation of a DBA or transient.
As discussed in Section 3.5 and summarized in Table 4-1 (item 294) of NEDO-31466, the IFTS was found to be a non-significant risk contributor to core damage frequency and offsite releases. IP has reviewed this evaluation, considers it . applicable to CPS, and concurs with the assessment.

== Conclusion:==

Since the screening criteria have not been satisfied, the IFTS LCO                                                l and Surveillances may be relocated to other plant controlled                                                     ;
documents outside the Technical Specifications.
CLINTON POWER STATION         Page 35 of 35 Appendix A: Justification
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