View source for U-603276, Forwards Rev 8 to CPS USAR, Which Includes Changes from 970701 Through 990630 & Available Changes Approved After 990630 That Reduce Overall Burden of Submittal Changes for Forthcoming Reporting period.10CFR50.59 Rept Included

{{Adams
| number = ML20217K516
| issue date = 10/08/1999
| title = Forwards Rev 8 to CPS USAR, Which Includes Changes from 970701 Through 990630 & Available Changes Approved After 990630 That Reduce Overall Burden of Submittal Changes for Forthcoming Reporting period.10CFR50.59 Rept Included
| author name = Coyle M
| author affiliation = ILLINOIS POWER CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000461
| license number = 
| contact person = 
| document report number = U-603276, NUDOCS 9910260182
| package number = ML20217K528
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 350
}}

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                                                                                                           . CanP,n Power Stato e
                                                                                                           ; P G Box 678
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l Chnton IL 61727 1 ei 217 935 8881
-                       ILLIN9BS P(MER                                                                           E4127 October                            8, 1999
 +
v Docket No. 50-461                                                                10CFR50.71(e)
Document Control Desk Nuclear Regulatory Commission 1                         Washington, D.C. 20555 t
i                   

==Subject:==
Clinton Power Station Submittal of the UDdated Safety Analysis ReDort. Revision 8

==Dear Madam or Sir:==

Pursuant to 10CFR50.71(e), Illinois Power (IP) hereby submits one original and ten copies of Revision 8 to the Clinton Power Station (CPS) Updated Safeiy Analysis F                          Report (USAR). The submittal indudes changes from July 1,1997 through June 30, 1999, as well as available changes approved after June 30,1999, that reduce the overall burden of submittal changes for the forthcoming reporting period In accordance with 10CFR50.59(b)(_), the attached 10CFR50.59 report (Attachment 2) identifies all safety evaluations for implemented changes. Yhis report
_                          includes those safety evaluations for changes affecting the USAR as required by 10CFR50.71(e)(2)(iii).
The revised USAR pages (Attachmant 3) are annotated with Revision 8. Filing 2                          instructions are included in Attr*nent 3. Although not specifically required by regulation, Attachment 4 proviou a summary of the changes made to the CPS Operational Requirements Manual since the submittal of the CPS USAR Revision 7 to the NRC, through June 30,1999.
.                                  As required by 10CFR50.71(e)(2)(i), the information given in the attachments accurately presents changes made since tne previous submittal and analyses submitted to the Commission or prepared pursuant to Commission requirement. Notwithstanding, IP qq'    acknowledges that CPS USAR discrepancies do exist. Known USAR discrepancies are t            '@'    ~
evaluated, trachd, and corrected via the CPS corrective action program A Current Licensing Batis poject is in progress at CPS. One aspect of this project is to identify and resolve USaR discrepancies in accordance with the NRC's policy on enforcement discretion for USAR discrepancies. This project is entering the discrepancy resolution
,                          phase. Discrepancy resolutions are being processed in accordance with approved station procedures.
                                                                                               /9 9910260182 991000                                                             V(h3,    ,
PDR K
ADOCK 05000461 PDR l   i dD-
                                                                                                     \
N

U-603276 Page7 Attachment 1 provides an affidavit supporting the facts set forth in this letter.
Sincerely yours, M.T bh M. T. Cople Assistant Vice President GBS/krk Attachments cc:    NRC Clinton Licensing Project Manager NRC Resident Office, V-690 Regional Administrator, Region III, USNRC Illinois Department of Nuclear Safety I
s
                                                                                                   .]

I      -
Attachment 1 l  .
to U-603276 Page1of1 Michael T. Ccyle, being first duly sworn, deposes and says: That he is Assistant Vice President at Clinton Power Station; that this letter supplying information for the Updated Safety Analysis Repod, Revision 8, has been prepared under his supervision and direction; l        that he knows the contents thereof; and that to the best of his knowledge and belief said l
letter and the facts contained therein are true and correct.
l Date: This 9w day of October 1999.
Signed:        U 7. 4 M. T. Coylg l
I 1
STATE OFILLINOIS                           l SS.                                                 I f
3, d##                  COUNTY             j j
l' Subscribed and sworn to before me this ? ^ day of October 1999, l
                                .m g.                       ,;
Joaquelhe S. NeitNes                            0..L 6                   g
                                                            ?l
                                                                                      ~
                  ;         Neary Putsc, men of Encis
:l uy em Egiros 11M2001 y                                    /p-' otary Public '
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Attachment 4 to U-603276 Page1of3 Operational Requirements Manual Changes The following revisions to the Clinton Power Station (CPS) Operational Requirements Manual (ORM) have been made in accordance with approved station procedures.
           - Changes to the ORM are reviewed per 10 CFR 50.59. These changes have been implemented since the submittal of the CPS USAR Revision 7 to the NRC, through June 30,1999.
           ' Revision Number _           ' Scope of Revision
           ' Revision 17
* Revised se', tion 6.5.2.7 to clearly define the Nuclear
                                      \        Review and Audit Group function associated with reviewing Independent Safety Engineering Group reports and the Facility Review Group reports.
* Revised Table 12, Attachment 2-19, Column Two heading by changing RESPONSE TIME to SETPOINT.
P 'sion 18                   e    Revised MSIV Leakage Control System Testing Requirements, items 4.4.7.2 and 4.4.7.3 (page 74) to remove testing level instrumentation. CPS does not have level instrumentation inputs to MSIV-LCS.                     I e    Revised page 79, sections 3.5.2 and 4.5.2.2 to pennit-
                                              . temporarily placing the motor operated valve thermal overloads in service during normal operation and routine plant operations, as well as for testing purposes. Guidance is provided in 3.5.2 time limitations.
Revision 19 '-               'e    Revised Facility Review Group Composition Section 6.5.1.2 (page 92) to reflect previously approved USAR change 6-70 and Safety Evaluation 94-051. That change package changed the title from Supervisor Nuclear Engineering to Senior Engineer-Nuclear.
* l Revised page 138, Attaciunent 3-5, to delete the open direction in the table for valves ISX014A, ISX014B, ISX014C.

l f                                                                                  Attachment 4 to U-603276 l                                                                                    Page 2 of 3,   _
Revision Number     Scope of Revision Revision 20'..      . Incorporate organizational changes made to section 6.0.        3
:e Revise the noun names for RHR valves IE12-F006A and                  i 1E12-F006B per Engineering Change Notice 30430.
* Revise the testing interval specified for snubbers from once J
per 18 months to once per refueling interval.-
i e . Change a Facility Review Group responsibility from reviewing the Licensee Event Reports to reviewing the reportable events.
Revision 21.        r. Incorporate organizational changes made to Sections 6.2.3.1 and 6.2.3.4. The change incorporates the               1 formation of the Nuclear Safety and Performance Improvement (NSPI) Department.
* Incorporate changes to section 6.5.2.10. The topic is the time frame to fonvard NRAG meeting minutes and reviews to the Vice President.
Revision 22 '       e    Revise the flow rate requirement for the control room ventilation system smoke mode. The editorial change removes "less than or equal to" from Testing Requirement '
4.4.3, leaving the flow requirements as "64,000 cfm 10%."
                                 .  . Delete the temperature parameter from the Channel Functional test requirements of Testing Requirement 4.4.7.2 for the MSIV Leakage Control System.                    )
* Delete staffing requirements from the ORM, refers users to      j USAR Section 13.1.2.3. This change was initiated by a           !
change to the Emergency Plan (ACN 12/2), and consolidates staffing requirements in the USAR.
Revision 23        'e    Revise Attachment I to show new Remote Shutdown                 i System Controls (Modi 6 cations IA-021, MS-044, RT-              l 039). Pages 109 through 111 were revised to allow adding line items on page 112.-
e    Revise ~ Attachment 2, Table 12 to reflect revised RPS Electric Power Monitoring Trip Setpoints.
l Revision 24 -   ,
e . Revise Attachment 4, Containment Isolation Valves, pages 148 and 154, to remove the '#' Applicable Mode from valves ISF001, ISF002,1B21-518A, and 1B21-518B.
o
* Attachment 4 to U.603276
           -                                                    - - -  _ _ . _..           Page 3 of 3 Revision Number     Scooe of Revisio_r1                                         j i
                     ,   ' Revision 25        e  Revise Attachment 1-4, Remote Shutdown System            1 Controls, to add components from modification FW-040
* Delete isolation requirement for condenser vacuum pump from table 3.2.16 -1, item 6.
* Delete item 2.4.3, Control Room Ventilation Smoke Mode.
* Revise Attachment 1-2 (p. I10) and Attachment 3-5 (p.
138) to delete valve ISX173A.
I e  Revise Attachment 3-1 (p.134) to remove valves CY020 and CYO21 per ECN 31400.
* Revise Attachment 3-5 (p.138) to delete valve ISX173B per ECN 31406.
* Revise Attachment 3-1 (p.134), Attachment 4-9 (p.148),
Attachment 4-10 (p.149), Attachment 4-14 (p.153),
Attachment 4-36 (p.175) for CC valves.                   ;
e  Revise Table 11 (p.129) per modification AP-048.         l (License Amendment 122) e  Revise Table 5 (p.120) to incorporate new trip setpoints for LPCS and LPCI A Pump Discharge Pressure ~ ligh, and Table 5 (p.121) for LPCI B and C pumps.
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%'                                           Attachment 2 to U-603276 Page 1 of 252 CLINTON POWER STATION 10CFR50.59 REPORT 1
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Attachment 2 to U-603276
* Page 2 of 252 REACTOR RECIRCULATION PUMP VIBRATION MONITORING Document Evaluated: Modification RRF015                                          Log Number: 93-04$ R/l          a 1
l Modification RRF015 makes permanent two temporary modifications (TMs) previously installed.                      l TM 90-031 reported for Updated Safety Analysis Report (USAR) Revision 3 installed a vibration monitoring system on the Reactor Recirculation (RR) system pumps. The monitoring system consists of two accelerometers and two velocity transducers for each pump. The accelerometers were mounted on the pumps' stuffing box lifling lugs using welded suppo t blocks. The accelerometers measure pump casing vibration. The velocity transducers are mounted on support                   i blocks welded to the top of the motor casing and measure lateral motor casing vibration. TM 92-007 later installed two horizontal proximity prches and axial proximity probes on each pump.                     ;
These are also to monitor pump vibration. RRF015 changes these two TMs to perman.% status                        j and installs a personal computer in the control room computer room to provide contittuous                        i monitoring and analysis of vibration data for the RR pumps and motors. This modification changes USAR Appendix E and F fire protection information. The installation was evaluated for                    i seismic loads, divisional separation, and fire loading and was found acceptable. This modification l
provides a monitoring function only; no trips or interlock permissives are associated with the                  j operation of the vibration monitors. No new failures or accident types are created by the change                j since the new equipment is passive in nature and seismic and fire protection concerns have been evaluated. This change has not been fully completed as of the cut-off date for USAR Revision 8.                ,
However, portions have been installed. An updated status of this change will be reported in Revision 9 to the USAR.
                                                                                                                   )
ACTUATOR AND SUPPORTING IIARDWARE REMOVAL FROM ECCS INJECTION                                                    l CHECK VALVES                                                                                                     l 1
Document Evaluated: Modification M-079                                   Log Number: 93-114 Modification M-079 removes the actuator hardware from the Emergency Core Cooling (ECCS)
System testable check valves. This change prevents the remote operations of valves 1E12F041 A, IE12F041B, IE12F041C, IE21F006, and IE22F005 during power operation. This remote operation capability was to satisfy a ponion of the ASME Code, Section IX. An ASME Code, Section XI, relief request (2014) was submitted and approved by the NRC on September 13, 1993, to allow full stroke exercising of these valves only during refueling operations. This will be done by the lever which will still be installed on the split shaft and will allow the valve to be tested locally. Therefore, the valves will still function as testable check valves. These valve.s are unassociated with any assumed accident initiators. They are associated with accident mitigatory equipment; however, their design basis fur. ion and operating characteristics are unchanged by l  this modification. Therefore, the probability and consequences of an accident or malfunction of equipment important to safety have not increased. This modification can not be postulated to initiate an accident or malfunction of equipment important to safety of a type difTerent than that previously evaluated. This rNdifiCation does not impact the Technical Specification requirements, nor does it impact the assumptions and bases used to establish those requirements.
t I
i j

s' Attachment 2 to U-603276 Page 3 of 252 Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification. This change has not been fully implemented as of the cutoff date for USAR Revision 8. However, portions have been installed. An updated status of this change will be reported in Revision 9 to the USAR. USAR text and figures are being revised as portions of this modification are released for operations.
ELIMINATION OF THE STEAM CONDENSING MODE OF RHR OPERATION Document Evaluated: Modification RH-033                               Log Number: 93-115 The steam condensing mode of the Residual Heat Removal (RH) System was intended to be used to increase plant availability by maintaining a hot standby condition until the plant could be restaned. This would be done by directing main steam through a Reactor Core Isolation Cooling (RI) System line into one of the two RH heat exchangers. The startup testing of this mode of operation was deleted by Safety Evaluation 87-1353 and reported in Updated Safety Analysis Report (US AR) Revision 0. As such, removing this as a startup test requirement and as a mode of operation at Clinton Power Station did not decrease the margin of safety or increase the consequences of an accident. In addition, Temporary Modification 90-04 removed the piping spool between the RI system piping and the RH heat exchangers. This temporary modification was reported for USAR Revision 2. Modification RH-033 makes this change permanent. This change will allow seven pipe hangers to be changed from snubbers to struts. The change provides for the de-termination of twelve Motor Operated Valves at the Motor Control Center and de-terminating of cables entering the Main Control Room. RH-033 includes extensive USAR changes to eliminate reference to this mode of RH operation. The use of the steam condensing mode including the capability to vent noncondensibles is not considered in any design basis           l accident. The isolated nature of the system can not function as the initiator of any evaluated        l accident or malfunction of any equipment important to safety. Nor can this isolated system be considered as a credible initiator of an accident or malfunction of equipment important to safety of a different type than those previously evaluated. The Technical Specifications were issued            J predicated on the absence of this cooling mode. Therefore, documentation of the abandonment of        )
the steam condensing mode of RHR operation does not result in a reduction of a margin of safety       j defined in the basis for any Technical Specification.                                                 l I
This change has not been fully complete as of the cutoff date for USAR Revision 8. However, portions have been installed. An updated status of this change will be reported in Revision 9 to      l the USAR. USAR text and figures are being revised as portions of this modification are released       l for operations.                                                                                       I MODIFICATION TO NSPS INVERTERS Document Evaluated: Modification IPF004                                       Log Number: 94-024     ,
Modification IP-F004 makes three changes to the Nuclear System Protection System (NSPS) inverters. The first change improves the system's capability to withstand DC system voltage 1

o
                                                                                           ' Attachment 2 m                                                                                        : to U-603276
  ,                                                                                        ,Page 4 of 252 transients by increasing the capacitance of the input filte capacitor and adding a blocking diode
         .' on the input bus. - The second change qualifies the system to operate at a low input voltage (less than 100 VDC) compared to the present trip setpoint (102 VDC). The third change installs a .
maintenance switch to allow the Division I and II inverters to be removed from service for inverter calibration and maintenance. This changeimproves the _overall reliability of the NSPS
         - power supply (including inverter fuses). The changes ensure that the system will function as                  -
designed during all analyzed events.
This change _ was reported in Revision 6 to the USAR. At that time, the change _ was partially completed. This change has not yet been fully complete as of the cutoff date for USAR Revision 8. An updated status of this changt will be reponed in Revision 9 to the USAR.'
RESOLUTION OF THE THERMO-LAG ISSUE FOR FIRE ZONE A-1 A Activity Evaluated: Modification FP-098, Supplement 1                        Log Number: 94-056 R/l
         - The original evaluation concerned the acceptability of Thermo-Lag fire wrap material that is installed to protect Division 2 power, control and instrumentation cables in Fire Zone A-1a even though the material did not fully provide a one-hour fire rating. As documented in Condition
       . Report.1-92-07-024-1, the Nuclear Regulatory Commission expressed concerns with the continued reliance on the Thermo-Lag without performing additional testing and evaluations.
Supplement 1 of Modification FP-098 includes the following activities: abandon Thermo-Lag in i place; construct a 3-hour'_ fire-rated 8" thick solid concrete block wall (including penetration seals) to create a new Fire Area A-6; provide three unlocked 3-hour fire-rated doors in the new wall for access between Fire Area A-6 and Fire Zone' A-la; add fire detectors in the new Fire Area, which serve to alert locally and to alert Operators in the Main Control Room; add normal, standby, and emergency lighting in the new Fire Area; replace stairs with a ladder to provide access to a
          . platform and flood door between the Auxiliary Building and Turbine Building basements; and blank off two_ duct penetrations in Fire Zone A-1a to upgrade the Fire Zone barriers to a 3-hour fire rating.
                                                            ~
The new fire wall and associated appurtenance will be designed and constructed in accordance
           -with existing design criteria and construction standards. The electrical load increase due to the addition of regular and emergency lighting and exit signs were evaluated and found to be within t he available capacity and will not have any adverse effect on the voltage. The impact of heat generated by power cables, cable trays, panels, and lighting was determined to be enveloped by the profile used for environmental testing of the safety-related components. Considering the new and revised configurations, normal operating temperature, humidity, and air flow pressure drop in
         - the areas were evaluated and found to be slightly increased, but were within previously defir.ed plant design limits. A new fire load calculation for the new Fire Awa A-6 establishes the fire load
          ' for the area to be moderate. Therefore, this activity does not increax the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). A postulated fire in Fire Zone A-la has been previously documented in the Fire Hazards Analysis which assures that one of the Safe Shutdown l
!                                                                                                                          J

o y                                                                                    Attachment 2 to U-603276 Page 5 of 252 Methods is available to support reactor shutdown during a fire scenario with an assumed loss of offsite power. The installation of a qualified fire wall to address the Divisional sepa;ation issue maintains a level of protection for safe shutdown equipment that is equal to or higher than that evaluated for.other plant areas. The potential impact of the presence of the wall on surrounding components has been evaluated and determined acceptable. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The Fire Protection Program is not included as part of the Technical Specifications. This activity does comply with the separation and protection requirements in Section III.G of 10CFR50, Appendix R, for safety-related equipment and cables.
All acceptance values or design limitations involving the new fire wall were previously documented for fire barriers of this type in USAR Appendix E and F. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
RESOLUTION OF THE THERMO-LAG ISSUE FOR FIRE ZONE CB-1e Activity Evaluated: Modification FP-099 and                                  Log Number: 94-074 R/l USAR Change 8-200 The original evaluation addressed the acceptability of Thermo-Lag in Fire Zone CB-le even though the material did not fully provide a one-hour fire rating. As documented in Condition Report 1-92-07-024-1, the Nuclear Regulatory Commission (NRC) expressed concerns with the continued reliance on the Thermo-Lag without performing additional testing and evaluations.
Modification FP-099 provides for the upgraded 1-hour rated Thermo-Lag fire barrier, with a provision for a partial coverage fire suppression system. In addition, the existing automatic sprinkler system is being modified to extend coverage to safety-related equipment. Although the defense-in-depth concept ofRevision 0 remains valid, this revision restares the use of Thermo-Lag with a 1-hour fire rating and modifies the fire suppression system in the area.
Calculations to support the upgrade of the Thermo-Lag included a revised combustible loading for the additional material and found the fire load to still be " low" In Fire Zone CB-le, Division 2 cable trays containing Method 2 safe shutdown cables are routed less than 20 feet from Division 1 cable trays containing Method 3 safe shutdown cables. T he concern is that a fire in the cable trays of either Division will spread to the other Division trays and damage the safe shutdown cables for both Safe Shutdown Methods. This is preverted by enclosing the Division 2 trays in 1-hour fire rated Thermo-Lag and protecting the cable trays of both Divisions with an automatic suppression system. This approach allows either Method 2 or 3 to be the Safe Shutdown Method for this fire zone.
The use of the upgraded Thermo-Lag as a qualified firebreak material is supported by a comparison and evaluation of critical parameters of the configuration tested by Nuclear Energy Insurers, meeting the requirements of NRC Generic Letter 86-10, Supplement 1. The use of the upgraded Thermo-Lag material as a qualified 1-hour firebreak, together with a fire detection and suppression system, meets the requirements of Section Ill.G.2.c of 10CFR50, Appendix R. The modification of the automatic sprinkler system in Fire Zone CB-le enhances the fire protection capability of the zone by the removal of sprinkler head obstructions and correcting sprinkler head

Atttchment 2 to U403276 Page 6 of 252 i area coverage. This activity reestablishes full comphance with the function, design, and
            - operational requirements previously established for these systems. Therefore, this activity does t not increase the probability or consequences of an accident or malfunction ofequipment important to safety previotisly evaluated in the Updated Safety Analysis Repon (USAR)._ A 7 postulated fire in Fire Zone CB-le has been previously documented in the Safe Shutdown -
Analysish The use of the qualified material and the modification of the existing sprinkler system        -
             " assures that one of the Safe Shutdown Methods is available to suppon reactor shutdown during a fire and assumed loss of offsite' power.. The modified fire suppression system and the upgrade of the Thermo-Lag material will be designed and constructed in accordance _with existing design criteria and construction standards and does not ' pduce any new failure mechanism to equipment surrounding it. Therefore, this activity does not create the possibility of an accident or l malfunction of equipment imponant to safety of a different type than previously evaluated in the USAR. The Fire Protection Program is not included as part of the Technical Specifications. This activity does comply with the separation and protection requirements in Section III.G of 10CFR50, Appendix R, for safety-related equipment and cables. All acceptance values or design limitations involving the new fire wall were previously documented for fire barriers of this type in USAR Appendix E and F. Therefore, this activity does not reduce a margin of safety as defined in the. basis for any Technical Specification.
CONDENSATE FILTRATION SYSTEM MODIFICATION Activity Evaluated: . Modification CP-020                                   Log Number: 94-075 R/l L Modification CP-020 changes the Condensate Cleanup System (CCS) by introducing the
            - Condensate Filtration (CF) system. The condensate filters will be placed upstream of the condensate demineralizers. One filter will be located upstream of each demineralizer internal to the existing demineralizer cubicle, The CF filters remove non-soluble impurities, primarily insoluble iron, from the condensate prior to the deep bed condensate demineralizers. Three cubicles were installed and released for operation. This modification does not impact plant operation because the CCS can operate using eight cells and maintain adequate flow and pressure
             - drop for the system. This modification also installs a flow control valve on the inlet of each filter / polisher train. The purpose of the flow control valve is to prevent excessive flow surges through each of the operating filter /demineralizer trains. Reducing the amount ofinsoluble iron entering the demineralizers improves the life and performance of resin beads, reduces the need for periodic condensate polisher cleaning, which results in reduced liquid and solid radwaste, and reduces activation of reactor coolant impurities.
The addition of the filters does not introduce a credible failure mechanism into the CCS system
             . beyond that already present. Failure modes of the filters, pressure boundary, flow control valve, and control system were evaluated and determined to not appreciably increase the probability of a
              ' Updated Safety Analysis Report (USAR) Section 15.6, Decrease in Reactor Coolant Inventory event. Credible failure modes of the filters and associated changes uhimately resu't in the same failure of the CCS or feedwater system as presently evaluated in the USAR., Therefore, the consequences of an accident previously evaluated in the USAR have not increased. These t

l l 3                                                                                   Attachment 2 to U-603276 Page 7 of 252 changes are associated with a non-safety related system and have been evaluated as not impacting equipment classified as imponant to safety. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The credible failure modes of the component and systems evaluated in this change remain bounded by the accidents evaluated in US AR Chapter 15, and the l    malfunctions of equipment important to safety evaluated in various sections of the USAR; and            -
therefore, this activity does not create the possibility of an accident or malfunction of equipment impcc t to safety of a different type than previously evaluated. This change does not affect any Techt:ical Specification corrponent and is un-related to supporting operability of any Technical Specification component. Derefore, this change does not reduce a margin of safety as defined in
!    the basis for any Technical Specifications.
INSTALLATION OF 3M INTERAM FIRE BREAK IN CONTAINMENT l
Activity Evaluated: Modification FP-097, Supplement I and                     Log Number: 94-076 R/:      l USAR Change 8-166 l                                                                                                              4 The original evaluation addressed the acceptability of Thermo-Lag in Containment Fire Area C-2 even though the material did not fully provide a one-hour fire rating. As documented in Condition Repon 1-92-07-024-1, the Nuclear Regulatory Commission (NRC) expressed concerns with the continued reliance on the Thermo-Lag without performing additional testing and evaluations. Modification FP-097, Supplement I replaces Thermo-Lag with 3M Interam material for ponions of three cable trays and associated hangers and conduit.
In Containment, above elevation 803 feet, Division 2 cable trays are routed within 6 feet of              1 Division 1 cable trays. The concern is that a fire in the cable trays of either Division will spread to   )
the other Division trays and damage the safe shutdown cables. This is prevented by placing a firebreak in the Division 2 trays. The use of 3M Interam as a qualified firebreak material is supported by testing ofNuclear Fire Protection Association Standard 251 meeting the requirements of NRC Generic Letter 86-10, Supplement 1. This activity does not reflect any                l change to the availability performance, testability, or reliability of the Fire Protection System and other plant hardware, systems, or procedures. The 3M Interam material was exposed to                      l cumulative radiation doses of 200 million rad with no degradation of the original physical properties. The installation meets applicable design and seismic considerations. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Repon
     ~(USAR), nor does it create an accident or malfunction of equipment important to safety of a difTgrent type than previously evaluated. The Fire Protection Program is not included as part of the Technical Specifications. However, this activity. complies with the separation and protection requirements in Section III.G of 10CFR50, Appendix R, for safety-related equipment and cables.
All acceptance values or design limitations involving the new fire barrier were previousic documented for fire barriers of this type in USAR Appendix E and F. Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.

I
* l Attachment 2            !
to U-603276 l
Page 8 of 252
{
I RESOLUTION OF THERMO-LAG ISSUE FOR FIRE ZONE CB-1f                                                       j Activity Evaluated: Modification FP-100 and                                   Log Number: 94-079 R/l USAR Change 8-148 The original evaluation addressed the acceptability of Thermo-Lag in Fire Zone CB-lf even though the material did not provide a one-hour fire rating. As documented in Condition Repon 1-92-07-024-1, the Nuclear Regulatory Commission expressed concerns with the continued reliance on the Thermo-Lag without performing additional testing and evaluations. Modification FP-100 provides for an upgraded 1-hour rated Thermo-Lag fire barrier, together with the installation of an automatic fire suppression sprinkler system covering a portion of the area.
Although the defense-in-depth concept of Revision 0 remains valid, this revision restores the use of Thermo-Lag with a 1-hour fire rating and installs a partial coverage automatic fire suppression system. Calculations to support the upgrade of the Thermo-Lag included a revised combustible loading for the additional material and found the fire load to still be " moderate" In Fire Zone CB-lf, a Division 2 cable riser containing Method 2 safe shutdown cables is routed within 19 feet of Division 1 cable trays containing Method I and 3 safe shutdown cables. The concern here is that a fire in the Division I cable tray will spread to the Division 2 trays and damage the safe shutdown cables. This modification encloses the Division 2 trays in 1-hour fire rated Thermo-Lag and protects the Division I cable trays with an automatic suppression system. This approach allows Method 2 to be the Safe Shutdown Method for this fire zone.
The addition of Thermo-Lag material, a combustible, was determined not to increase the fire load classification. This activity enhances the Divisional separation of the Safe Shutdown Methods.           i This activity reestablishes full compliance with the function, design, and operational requirements previously established for these systems. Therefore, this activity does not increase the probability     {
or consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the Updated Safety Analysis Repon (USAR). A postulated fire in Fire Zone CB-lf
{
1 has been previously documented in the Safe Shutdown Analysis. The use of the qualified material and the addition of the sprinkler system assures that Safe Shutdown Method 2 is available to suppon reactor shutdown during a fire and assumed loss of offsite power. The new fire suppression capability and the upgrade of the Thermo-Lag material will be designed and         ~
constructed in accordance with existing design criteria and construction standards and does not          q introduce any new failure mechanism to equipment surrounding it. Therefore, this activity does not create the possibility of an accident or malfunction of equipment imponant to safety of a different type than previously evaluated in the USAR. The Fire Protection Program is not included as part of the Technical Specifications. However, this activity complies with the separation and protection requirements in Section III.G of 10CFR50, Appendix R, for                      I safety-related equipment and cables. All acceptance values or design limitations involving the new fire barrier were previously documented for fire barriers of this type in USAR Appendix E            j and F. Therefore, this activity does not reduce a margin of safety as defined in the basis for any       j Technical Specification.                                                                                 I 1
1 1
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1 Attachment 2 to U-603276 Page 9 of 252 DISCONTINUING SOME WATER CHEMISTRY MEASUREMENTS FOR CONDENSATE                                        I AND FEEDWATER Document Evaluated: USAR Change 8-070                                         Log Number: 95-082 (ref. ECN 28678)
Engineering Change Notice (ECN) 28678 removed turbidity cells 1 AE-PS026/029 and silica analyzers I AIT-PS024/025 from panel IPL88JA. Changes were made to Updated Safety l
Analysis Report (USAR) Figure 9.3-4 in accordance with ECN 28678. This change removes the footnotes at the end of the design bases for the condensate cleanup system output in USAR section 10.4.6.1. Measurements are not typically taken for silica, chloride, and sodium.
However, the footnotes state that cenain parameters are not measured when the conductivity is within certain limits. This change removes the silica analyzers and the turbidity cells W performing measurements implied by section 10.4.6.1 for the condensate and feedwater sysems.
Assuming that not monitoring the chemistry parameters being deleted by this change could allow detrimental . water in the loop, the chemistry measurements taken for the reactor water would detect a problem and allow for corrective action. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety evaluated -
previously in the USAR or of a different type. Poor water chemistry could cause corrosion or plating of deposits on components. The condensate and feedwater systems are not accident              ,
mitigating systems; and the water chemistry for accident mitigating systems is monitored by the measurements taken for the reactor water. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the US AR or of a different type. Conductivity measurements or acceptance standards are not changed by this revision.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
CARPET TILE INSTALLATION IN MCR, TSC AND OSC                                                         l Activity Evaluated: ECN 29418                                                 Log Number: 95-092     )
l Engineering Change Notice (ECN) 29418 added tile type carpet in the Main Control Room                l (MCR), the Operations Support Center (OSC), and the Technical Support Center (TSC).                  j Updated Safety Analysis Report (USAR) Appendix E, " Fire Protection Evaluation Report," was          j revised to account for the new fire loading. Vendor documentation was reviewed to determine          !
the flame spread, smoke, fuel contribution, as well as for any adverse material interaction of the   l carpet. NUREG-0800, " Standard Review Plan," was reviewed to determine the effects of any            '
smoke or heat on the area ventilation systems and to determine the electrostatic effect of the carpet on the control room instrumentation. The addition of the carpet has been determined to        l I
have no adverse effect on the equ'ipment, instmmentation, or ventilation in the MCR, TSC, and OSC. The carpeting consists of tile which are nylon-faced and a backing which consists of polyester, vinyl, and calcium carbonate. The addition of the carpeting increases the fire loading in fire zones CB-6a, CB-6b, CB-6c, and CB-6d; however, even with the increase, the fire severity

4     j Attachment 2 to U-603276                 ,
Page 10 of 252 remains " moderate" in all four zones, Any smoke produced will be readily detected by smoke detectors in thc MCR area and the MCR Heating, Ventilating, and Air Conditioning System (VC) ducts which will automatically initiate MCR annunciation and the VC smoke isolation mode to recirculate the MCR air through carbon filters. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR, nor does this activity create the possibility of an accident or        -
malfunction of equipment important to safety of a different type than previously eviuated. This activity does not adversely impact any Technical Specification safety limit, limiting safety system setting, or limiting condition for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
ACTUATOR GEARS CHANGE FOR MSIV LCS BLEED VALVES Activity Evaluatedi ECNs 29241 & 29242                                         Log Number: 95-097 Engineering Change Notices (ECNs) 29241 and 29242 modify the gear ratios in the Limitorque Actuators for Main Steam Isolation Valve (MSIV) Leakage Control System (LCS) Bleed Valves lE32-F006 and IE32-F009. This action increases opening and closure torque to ensure the valves' ability to perform their design function. Updated Safety Analysis Report (USAR)
Revision 7 reported a partial completion for this Safety Evaluation pertaining to the change for MSIV LCS Bleed Valve IE32-F006. Since USAR Revision 7 was issued, the modification for MSIV LCS Bleed Valve 1E32-F009 has been installed. This modification changes the Limitorque Actuator gear ratio from 33.5:1 to 43.75:1 for valve IE32-F009 which increases torque, but slows the valve closure rate. No circuit or interlock changes are associated with this modification.
The valve continues to operate and function in accordance with design requirements. These components are not assumed initiators of any evaluated accident and they do not function to prevent the occurrence of any evaluated accident. The valves and their associated system function in an accident mitigatory role. This change increases the assuredness that these valves can perform this mitigatory function. The potential effect of the increase in valve stroke time from 38 seconds to 48 seconds has been evaluated and determined to have an inconsequential effect on the radiological dose consequences of USAR evaluated accidents (bounded by the loss of coolant accident) which postulate leakage via this path. Therefore, this change does not increase the probability or consequences of a previously evaluated accident. The change in gear ratios for these actuators has not introduced any new failure mechanisms beyond those already present in the actuator design. This change does not directly or indirectly affect the function of any nearby equipment important, nor does it introduce any consequential failure of components within its l
associated system. Should a postulated malfunction or failure of this or a related system                 j component occur, the radiological consequences of the malfunction or failure would be unchanged from the same failure prior to this modification. Hence, there is no increase in the probability of a malfunction or consequences of a malfunction of equipment important to safety.
This modification has not introduced any failure mechanisms not already present in the original actuator design. Thus, this modification has not created a potential new type of accident or malfunction ofequipment important to safety. The Technical Specification requirements for                 ,
MSIV functionality and LCS ftmetionality are unchanged. The slight difference in valve stroke             i 1
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  .                                                                                   Attachment 2 to U-603276 Page 1I of 252 time has an inconsequential effect on accident consequences in which the MSIV LCS is considered to provide a mitigatory function. Therefore, this modification does not result in a reduction of a margin of safety as defined in the basis for any Technical Specification.
MOVE THE BACKWASH RECEWING TANK INLET CHECK VAL VE IG36F320 Document Evaluated: Modification RTF039                                      Log Number: 96-00lR/l Modification RTF039 moves the Backwash Receiving Tank (BWRT) inlet check valve IG36F320 to a new location ~ upstream of the Reactor Water Cleanup (RWCU) filter /demineralizer input lines to the BWRT. This eliminates a flow path which allows RWCU water from the filter /demineralizers to escape to Containment atmosphere through the RWCU Precoat tank lid and BWRT Cycled Condensate (CY) fill funnel if the inlet check valve sticks shut. That occurrence results in significant degradation of radiological conditions. The RWCU system does not perform any safety functions, except portions of the system provide an American Society of Mechanical Engineers (ASME) Section III pressure boundary. The system is not an assumed initiator of any evaluated accident. There is no safety related equipment in the immediate vicinity of the piping that could be damaged by vibration of the piping during a seismic event. Moving the valve to the new location does not adversely impact the integrity of the piping or its supports.
The new location of the check valve improves the system function by preventing backflow of the BWRT filter /demineralizer inlets to the precoat tank and BWRT funnel. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). This modification does not create failure modes or alter failure consequences beyond those also present in system design. Therefore, this change does not create the possibility of a different type of accident or malfunction of equipment iinportant to safety than previously evaluated in the USAR.
None of the requirements discussed in the Technical Specifications are directly affected by the RWCU filtering system. Nor does this activity affect the performance of mitigatory features used to ensure the integrity of the fission product boundaries. Therefore, this change does not reduce a margin of safety as defined in the basis for any Technical Specification.
MAIN STEAM ISOLATION VALVE PILOT AIR CONTROL SOLENOID VALVE REPLACEMENT MODIFICATION Activity Evaluated: Modification MSF031, Partial 2                          Log Number: 96-008 R/l This Safety Evaluation revises Safety Evaluation 96-008. Safety Evaluation 96-008 indicated that Partial 2 would be installed in Refueling Outage Number 7; however, due to problems with the outboard Main Steam Isolation Valves (MSIVs), Partial 2 was installed prior to startup from Refueling Outage Number 6.
This modification provides a pair of Automatic Valve Company single coil solenoid operated valves (SOVs) as a substitute for the discontinued ASCO dual coil nuclear SOV (ASCO series NP8323).
These valves control pilot air on the MSIV actuators. The mounting, air tubing and electrical l
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                                                                                   \                   4 Attachment 2 to U-603276 Page 12 of 252 l
connector are ofIllinois Power (IP) design. The design incorporates a multi-pin connector at the conduit entrance to the MSIV electrical box. This provides a new environmental seal to the conduit.        1 This is required due to the difference in qualification between the old and the new SOVs. The two          l new 3-way SOVs have been tubed to provide the same function as the old dual coil SOVs. The                 l' tubing schematic and the two individual solenoid coils maintain the design basis control logic that if both the "A" solenoid and the "B" solenoid are de-energized, the MSIV will close with both air and spring power. If only one solenoid is de-energized, the MSIV will remain open.
The probability of MSIV closures is a consideration for this modification. Since, the two coil system with each coil energized by a separate bus is maintained and the ability of either energized coil to keep the MSIV open is maintained, there is no change to the probability of a spurious closure challenging safety systems. This modification maintains the design, material, and construction standards for the MSIV. This modification does not affect the separate energy sources or the redundancy of the valves. The reliability of the valves in being able to effect MSIV closure has been evaluated and determined to be equivalent to the existing valves. It does not affect the stored air or spring energy used to effect MSIV closure. The seismic qualification for this modification was reviewed and the original qualifications and evaluations of the MSIV were maintained. This change does not affect the MSIV closure times as analyzed in the Updated Safety Analysis Report (USAR). Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. Since this modification maintains the function of the MSIV within the current design bases, it does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. Further, this modification is consistent with Technical Specification requirements and the bases for those requirements. Since this modification does not affect closure times or induce reactor transients, it does not reduce a margin of safety as defined in the basis for any Technical Specification.
OSCILLATION POWER RANGE MONITOR Document Evaluated: Modification NR-010 and                                    Log Number: 96-016 USAR Chang ~e 8-022 Modification NR-010 installed the Oscillation Power Range System (OPRM) into the Neutron Monitor System panels to meet the requirements of Generic Letter 94-02 and General Design Criteria 10 and 12. The OPRM system will monitor Local Power Range Monitors (LPRMs) and provide alarms and trips on the onset of thermal hydraulic oscillation transients. The Average Power Range Monitor (APRM) Fixed Neutron Flux-High Function is capable of generating a trip signal to prevent fuel damage or excessive reactor pressure. The OPRM equipment has been designed and installed to not degrade the existing APRM, LPRM, and Reactor Protection System (RPS) systems. These systems will still perform all of their intended functions. The new equipment was tested and installed to the same or better environmental and seismic envelopes as the existing systems. The OPRM equipment has been designed to single failure criteria and is electrically isolated from equipment of different electrical divisions or non-Class IE equipment.
The electrical loading is within the capability of the existing power sources and the heat loads are

.                                                                                   Attachment 2 to U-603276 Page 13 of 252 within the capability of existing cooling systems. The APRM and RPS systems are designed to fail in a tripped condition. To provide assurance that a minimum number of APRMs are OPERABLE, inoperability of the APRM also produces a trip condition. Therefore, this activity does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). There are no new failure modes of the either the new OPRM equipment or the existing APRM equipment. The new OPRM equipment is designed and installed to the same requirements as the existing APRM equipment and is designed and tested to have no impact on the existing functions of the APRM system. Therefore, this activity does not create the possibility of an accident or a malfunction of equipment important to safety of a different type than previously evaluated in the USAR. This activity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
UNIT 2 SOLID RADWASTE FP/FD SLUDGE TANK LEVEL INSTRUMENTATION Document Evaluated: Modification WX-030 and                                    Log Number: 96-018 USAR Change 8-024 Modification WX-030 replaces the ultrasonic continuous level monitoring instrumentation for Unit 2 Solid Radwaste (WX) Fuel Pool Cooling and Cleanup Filter Demineralizer Sludge Tank,
    ' 2WX02T, with on demand plumb bob type level measuring instrumentation to measure both total tank level and sludge level. Also, this modification installs a " Thermal Dispersion" Resistance Temperature Detector (RTD) type sensor element and circuitry for a continuous high-high level alarm. In addition, a pump trip and high and low sludge level alarms are being removed due to the manual nature of the new level monitoring system. The pump trip was eliminated because the plumb bob can not be used while the pump is running or the system is likely to be damaged. The sludge' level alarms are operator aids only; and could become more of a nuisance than help, because they only annunciate after the operator has manually requested tank level. Modification WX-030 is being made to improve the reliability of the level monitoring instrumentation for the tank _ Updated Safety Analysis Report (USAR) Figure 11.4-3, Sheet 3, Appendix E Figures FP-16a & b, Appendix E Cable Tray Figure 14, and Sections 7.7.1.12.3.1 are being revised as a result of this modification.
The only associated safety related failure evaluated in the USAR is the WX tank rupture. This new monitoring system does not increase the probability or consequences of that event, because this modification will result in providing more reliable information to the operator which should thereby reduce system challenges. The USAR does not address failure of WX tank level instrumentation. Therefore, this activity does not increase the probability or consequences of an    I accident or malfunction of equipment important to safety previously evaluated in the US AR. The plum bobs introduce a new component failure that is not a nuclear safety related concern, but has As-Low-As-Reasonably-Achievable (ALARA) consequences. To mitigate the possibility of an ALARA concern, preventive maintenance tasks are to be added that will replace the plumb bob's cable every 5 years, which is very conservative per the manufacturer. Therefore, this activity does i

4 Attachment 2 to U-603276 Page 14 of 252 not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This activity does not impact any components subject to the requirements of the Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation. Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
SAFE SHUTDOWN RECLASSIFICATION OF THE WHITTAKER CABLE IN FIRE ZONE CB-lF Document Evaluated: USAR Change 8-253                                           Log Number: 96-019 R/l As defined in Updated Safety Analysis Report (USAR) Appendix E, Fire Hazards Analysis, Fire Zone CB-1f contains safety-related equipment consisting of Division 1 and 2 cables. A Division 2 cable in this zone is the Whittaker cable IRP02Q, which was installed based on its 3-hour fire rating. This activity reclassifies this cable as not required for safe shutdown for a fire in Fire Zone CB-If. thereby removing the requirement for it to have a 3-hour fire rating. Since there are no             f physical changes to the cable and no other fire hazards are changed or added by this activity, there        l is no change to the combustible loading. In addition, there is no affect on the fire protection features of this zone, i.e., fire barriers, fire-rated penetrations, etc., that prevent the fire from spreading to another plant area. The Engineering Evaluation for Nuclear System Protection System (NSPS) Bus IC71-P001B determined that the Whittaker cable is not required to remain available to supply power to any safe shutdown loads for a fire in Fire Zone CB-lf, and therefore, the loss of the Whittaker cable does not affect Safe Shutdown Method 2. This allows Method 2 to continue to be a Safe Shutdown Method for this fire zone. There is no physical change to the cable or its availability, performance, and reliability are not adversely affected for its normal function. This activity does not increase challenges to safety systems or otherwise affect any structure, system, or component. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safay previously evaluated in the USAR. The reclassification of the Whittaker cable IRP02Q as a non-safety related Shutdown cable without the 3-hour fire barrier feature maintains the existing Safe Shutdown Method 2 previously evaluated as available to support reactor shutdown during a fire and assumed loss of offsite power. As such, it does not reflect any change to or add any credible failure modes associated with a postulated fire in Fire Zone CB-1finvolving the Fire Protection System and other plant hardware. ::ystems, or procedures. The loss of this cable and the associated NSPS bus due to a fire in the zone does not increase the severity of a fire in Fire Zone CB-lf. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The                 ,
Fire Protection Program is not included as part of the Technical Specifications. However, this              j activity complies with the separation and protection requirements in Section Ill.G of 10CFR50,              l Appendix R, for safety-related equipment and cables. Therefore, this activity does not reduce a             i margin of safety as defined in the basis for any Technical Specification.
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Attachment 2 to U-603276 Page 15 of 252 SOLID RADWASTE CONCENTRATE TANK B LEVEL INSTRUMENTATION Document Evaluated: Modification WX-031 and                                       Log Number: 96-028   ;
USAR Change 8-025 Modification WX-031 replaces the ultrasonic continuous level monitoring instrumentation for the        )
Solid Radwaste (WX) Concentrate Tank "B", 0WX0lTB, with on demand plumb bob type level measuririg instrumentation to measure total tank level. Also, this modification installs a " Thermal Dispersion" Resistance Temperature Detector (RTD) type sensor element and circuitry for a              3 continuous high-high level alarm. In addition, a pump trip and various level alaims are being          j removed due to the manual nature of the new level monitoring system. The pump trip was eliminated because the plumb bob can not be used while the pump is running or the system is likely to be damaged. Finally, an instrument air line with isolation valve and pressure regulator is being added to provide a positive pressure environment for the level sensor housing. The nominal increase in air flow from the level sensor housing to the tank has no appreciable effect on airborne radiation. Modification WX-031 is being made to improve the reliability of the level monitoring         ,
instrumentation for the tank. Updated Safety Analysis Report (USAR) Figue 3.6-1, Sheet 82,             j Figure 9.3-2, Sheet 6, Figure 11.4-3, Sheet 4, Appendix E Figures FP-19a & b, Appendix E Cable Tray Figure 17, and Sections 7.7.1.12.3.1 are being revised as a result of this modification The only associated safety related failure evaluated in the USAR is the WX tank rupture. This new monitoring system does not increase the probability or consequences of that event, because this modification will result in providing more reliable information to the operator which should thereby reduce system challenges. The USAR does not address failure of WX tank level instrumentation. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The plum bobs introduce a new component failure that is not a nuclear safety related concern, but has As-Low-As-Reasonably-Achievable (ALARA) consequences. To mitigate the possibility of an ALARA concern, preventive maintenance tasks are to be added that will replace the plumb bob's cable every 5 years, which is very conservative per the manufacturer. Therefore, this activity does    j not create the possibility of an accident or malfunction of equipment important to safety of a         l different type than previously evaluated. This activity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.                                                                                         l UNIT I SOLID RADWASTE PHASE SEPARATOR TANK LEVEL INSTRUMENTATION Document Evaluated: Modification WX-032 and                                       Log Number: 96-029 l
USAR Change 8-026 l
Modification WX-032 replaces the ultrasonic continuous level monitoring instrumentation for Unit 1 Solid Radwaste (WX) Phase Separator Tank,1WX03T, with on demand plumb bob type level measuring instrumentation to measure both total tank level and sludge level. Also, this

4 Attachment 2 to U-603276 Page 16 of 252 modification installs a " Thermal Dioersion" Resistance Temperature Detector (RTD) type sensor element and circuitry for a continuous high-high level alarm. In addition, a pump trip and high and low sludge level alarms are being removed due to the manual nature of the new level monitoring system. The pump trip was climinated because the plumb bob can not be used while the pump is running or the system is likely to be damaged. 'Ihe sludge level alarms are operator aids only; and could become niore of a nuisance than help, because they only annunciate after the operater has manually requested tank level. Modification WX-032 is being made to improve the reliability of the level monitoring instrumentation for the tank. Updated Safety Analysis Report (USAR) Figcre 11.4-3, Sheet 2, Appendix E Figures FP-16a & b, Appendix E Cable Tray Figure 14, and Sections 7.7.1.12.3.1 are being revised as a result of this modification.
The only associated satey related failure evaluated in the USAR is the WX tank rupture. This new monitoring system does not increase the probability or consequences of that event, because this modification will result in providing more reliable information to the operator which should thereby reduce system challenges. The USAR does not address failure of WX tank level instrumentation. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The plum bobs introduce a new component failure that is not a nuclear safety related concern, but has As-Low- As-Reasonably-Achievable (ALARA) consequences. To mitigate the possibility of an ALARA concern, preventive maintenance tasks are to be added that will replace the plumb bob's cable every 5 years, which is very conservative per the manufacturer. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This activity does not impact any components subject to the requirements of the Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
UNIT 1 SOLID RADWASTE FP/FD SLUDGE TANK LEVEL INSTRUMENTATION Document Evaluated: Modification WX-033 and                                    Log Numbcr: 96-030 USAR Change 8-029 Modification WX-033 replaces the ultrasonic continuous level monitoring instrumentation for Unit 1 Solid Radwaste (WX) Fuel Pool Cooling and Cleanup Filter Demineralizer Sludge Tank, lWX02T, with on demand plumb bob type level measuring instrumentation to measure both total            !
tank level and sludge level. Also, this modification installs a " Thermal Dispersion" Resistance Temperature Detector (RTD) type sensor element and circuitry for a continuous high-high level alarm. In addition, a pump trip and high and low sludge level alarms are being removed due to the manual nature of the new level monitoring system. The pump trip was eliminated becauce the plumb bob can not be used while the pump is running or the system is likely to be damaged. The sludge level alarms are operator aids only; and could become more of a nuisance than help,              l because they only annunciate after the operator has manually requested tank level. Modification WX-030 is being made to improve the reliability of the level monitoring instmmentation for the tank. Updated Safety Analysis Report (USAR) Figure 11.4-3, Sheet 3, Appendix E Figures

I Attachment 2 to U-603276 Page 17 of 252 FP-16a & b, Appendix E Cable Tray Figure 14, and Sections 7.7.1.12.3.1 are being revised as a result of this modification.
The only associated safety related failure evaluated in the USAR is the WX tank rupture. This new monitoring system does not increase the probability or consequences of that event, because this modification whl result in providing more reliable information to the operator which should      -
thereby reduce system challenges. The USAR does not address failure of WX tank level instrumentation. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The plum bobs introduce a new component failure that is not a nuclear safety related concern, but has As-Low-As-Reasonably-Achievable (ALARA) consequences. To mitigate the possibility of an ALARA concern, preventive maintenance tasks are to be added that will replace the plumb bob's cable every 5 years, which is very conservative per the manufacturer. Therefore. this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This activity does not impact any components subject to the requirements of the Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
HYDROGEN SYSTEM SAMPLING MODIFICATION Document Evaluated: ECN 29666                                                   Log Number: 96-043 Engineering Change Notice (ECN) 29666 installs 1/4 inch brass ball valves in place of pipe caps in the Hydrogen System panel 1PL10J. This ECN makes it easier for maintenance personnel to connect equipment for thermal gas analyzer calibration. The Hydrogen System continues to function as before without any change, therefore, this change does not increase the possibility or consequences of an equipment malfuncticn or an accident previously evaluated in the Updated Safety Analysis Report (USAR). The addition of this valve helps maintenance personnel and increases safety by eliminating the potential for explosive mixture in the atmosphere. The valve remains closed and the end capped during operation. Therefore, this change does not create the possibility of an accident or malfunction of equipment of a difTerent type than previously evaluated in the USAR. The Technical Specifications do not discuss the generator Hydrogen System, therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specifications.
PROVISION TO USE PLANT SERVICE WATER TO FLUSH CONDENSER TUBES Document Evaluated: CPS No. 3212.01                                      Log Number: 96-070 This change to Clinton Power Station procedure 3212.01 provides instructions for connecting temporary hoses downstream of Plant Service Water (WS) system valves 1WS050A/B and 1WS056A/B on the inlet and outlet lines to the Turbine Building Closed Cooling Water (WT) heat exchangers, lWT01 AA and IWT01 AB. These hoses will provide a source of water to flush I                     .

Attachment 2 to U-603276                    i Page 18 of 252
                                                                                                               ]
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out the main condenser tubes and/or waterboxes. Normally, the water is routed to pressure increasing flushing pumps to provide a source of high pressure water to flush silt and debris out of the condenser tubes and off orthe tubesheets. Connection of hoses and drawing water from these               l lines constitutes a change to the operation of the WS system as described in the Updated Safety              l Analysis Report (USAR).                                                                                      !
The amount of WS water to be used is very small in comparison to the capacity of the WS system; therefore, using WS water to flush the condenser tubes does not adversely affect the WS or WT system operations. The WS system and components are not required for safe shutdown of the plant and are not classified as important to safety. Therefore, this change does not increase the probability or consequences of a malfunction of equipment previously evaluated in the USAR.
The USAR discusses a flooding analysis for the turbine building, but the leak due to a rupture of the temporary hoses would be minuscule in comparison to the assumptions of this analysis.
Therefore, the connection of hoses and the use of water from WS in the turbine building 712' elevation does not increar, the probability or consequences of an accident previously evaluated in the USAR. This function will only be used when the plant is shutdown; thus, no failure could cause a plant shutdown or challenge any safety systems. The only possible malfunction scenario caused by this change is flooding, and it has already been analyzed. This change does not create the possibility of an accident or malfunction of equipment important to safety of a different type.
This change does not alter the assumptions used in deriving the Technical Specifications.
Therefore. this change does not reduce a margin cf safety as defined in the basis for any Technical Specification.
GENERATOR STATOR WATER COOLING AIR INJECTION Document Evaluated: ECN 29644                                                  Log Number: 96-094 Engineering Change Notice (ECN) 29644 installs 3/8" copper tubing, needle valves and a flow meter to allow injection ofInstrument Air into the Main Generator Stator Cooling Water (GC) system. Currently, the GC dissolved oxygen content ranges from 2.0 ppm to 0.03 ppm. The equipment manufacturer recommends that the dissolved oxygen content be maintained in the range of 2 to 8 ppm. This change provides a mechanism to increase dissolved oxygen content in the stator cooling water to the recommended value. By maintaining the recommended value of dissolved oxygen, the reliability of the generator is enhanced.
According to the Updated Safety Analysis Report (USAR), the GC system is classified as non-safety related. A malfunction resulting from this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The follow' m   g malfunctions of equipment could result from this change: (1) increase in the conductivity of GC water, (2) increase the probability oflecl:s in the stator winding, or (3) cause a runback of the generator and result in loss of turbine load. Eventually, these malfunctions will result in a loss of the main generator which could lead to a plant scram. Although the equipment malfunctions just discussed potentially initiate USAR evaluated accidents, such as a loss of generator load or turbine trip, this activity does not cause an increase in probability beyond 1

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    .                                                                                  Attachment 2 to U-603276 Page 19 of 252 that originally assumed in licensing the facility. Should a component failure occur, the resulting event will occur as assumed in the USAR with no resultant increase in offsite dose consequences.
Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The GC system only affects the main generator;it does not affect any system important to safety; and postulated failure result in already evaluated accidents.
Therefore, this change will not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The operational characteristics of the GC system are not associated with the establishment of any margin of safety defined in the basis for any Technical Specification Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
NEW CPS PROCEDU'RE 1003.02," FIELD CONFIGURATION CHANGE PROCESS" Document Evaluated: CPS Procedure 1003.02                                      Log Number: 97-025 This activity supports issuance of a new Clinton Power Station Procedure 1003.02, " Field Configuration Change (FCC) Process" and as ociated form, 1003.02F001. The new procedure allows qualified personnel to change the configuration of non-Quality related components without processing a design change document.
The new procedure continues to require existing controls on any change that affects a design or licensing basis function of the SSC. Changes to quality-related or ASME Code components or changes that affect equipment design, licensing basis, calculations, setpoints or require issuance of a new procedure or full procedure revision are not allowed by this new process. Due to these limitations, the configuration changes will not result in an increase in the probability or consequences of any accident or a malfunction of equipment important to safety evaluated previously in the USAR. Changes outside this scope, and changes to equipment within the scope of the QA Manual, will continue to be controlled under the existing design controls. Also, qualified individuals are still required to perform the configuration changes. The limits on the scope of the program provide assurarce that changes that involve an unreviewed safety question or atfect nuclear safety will continue to be reviewed by the Facility Review Group. Therefore, this activity will not create the potential for an accident or equipment malfunction of a difTerent type than previously evaluated in the USAR. The allowable configuration changes will not affect design margin or any of margin of safety. Therefore, issuance of this new procedure does not reduce a margin of safety as defined in the basis for any Technical Specification.
USAR INCONSISTENCIES REGARDING POSTULATED PIPE BREAKS Document Evaluated: US AR Change 8-223                                 Log Nember: 97-037 This activity revises inconsistencies in Updated Safety Analysis Report (USAR) Sections D3.6.2.2.1, D3.6.2.2.4, and D3.6.2.7.5 and Tables B3.6-18 and B3.6-19 regarding pcstulated pipe breaks. USAR Figures B3.6-3, B3.6-18 and B3.6-19 accurately show the postulated pipe breaks in the Auxiliary Building Steam Tunnel and the Residual Heat Removal (RHR) system heat t
w

                                                                                                            '4 Attachment 2 to U-603276 Page 20 of 252 exchanger rooms. These are not new or additional postulated pipe breaks, only a correction in
   . the USAR for inconsistencies. Text discussion has been added to Sections D3.6.2.2.1 and                      ,
D3.6.2.2.4 for the postulated terminal end breaks in the 10-inch RHR line and lines which connect l
into the 4-inch RHR lines. Section D3.6.2.7.5 has been revised to note that breaks are not                  j postulated in the RHR heat exchanger rooms. The postulated pipe breaks RH-C5 and RH-C315                    I have been deleted from Tables B3.6-18 and B3.6-19. These changes will provide consistency                    l with USAR Figures B3.6-3, B3.6-18, and 33.6-19.                                                             {
The proposed activity does not involve any physical changes, it merely clarifies the USAR for inconsistencies related to postulated pipe breaks. Therefore, the pro, nosed change does not increase the probability or the consequences of an accident previously evaluated in the USAR.                ,
Since this activity does not alter the Feedwater or Residual Heat Removal systems, it will not increase the probability or consequences of a malfunction of equipment important to safety evaluated in the USAR. Because the USAR already considered design basis flooding, environmental, pipe whip and jet impingement hazards analysis, and the sub-compartment pressurization analysis, there is no possibility of an accident of a different type than any evaluated previously. Since this activity does not change any plant components, this activity will not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the USAR. Therefore, this activity does not reduce a margin of safety margin of safety.as defined in the basis for any Technical Specification.
ELIMINATION OF THE STEAM CONDENSING MODE OF RHR Document Evaluated: Modification RH-033, Supplement 3,                           Log Number: 97-040 and USAR Change Package 8-032 4
Valves IE12-F073A&B were motor operated, normally closed valves which were designed to throttle open to vent steam and non-condensable gases from the Residual Heat Removal (RHR)
Heat Exchangers A and B while in or following the Steam Condensing Mode of RHR operation.                    1 Modification RH-033, Supplement I took the motor operators out of service, therefore, these valves are now manually operated globe valves. Since these valves were initially installed to vent reactor non-condensable gases and some steam during the steam condensing mode of RHR operation, they discharge below the normal water level of the weir in the Drywell. This would allow for the condensation of any steam and would route non-condcasable cases into the Drywell to avoid radiological challenges to the outer containment atmosphere while venting. RH-033, Supplement 1, Safety Evaluation Log # 96-026, stated that the valves IE12-F073 A&B had no function. The safety function of these valves as Drywell isolation valves was missed in Supplement i Valves IE12-F073 A&B are now considered Drywell boundary valves since during                    l a LOCA event, the water in the Drywell weir is blown out through the vent openings into the                  I outer containment suppression pool. If valves lE12-F073 A&B were left open, a potential steam bypass of the suppression pool exist through these valves during the above scenario. This Safety Evaluation corrects statements made in Safety Evaluation 96-026 and acts as corrective action for Condition Report 1-97-01-132. The following figures in the Updated Safety Analysis Report i
i

Attachment 2         l to U-603276          i Page 21 of 252       )
l (USAR) are being revised to show closed valves (IE12-F073 A&B) as locked closed:
Figure 3.6-1, sheets 62 and 63, Figure 5.4-13, sheets 1 and 2, and Figure 6.2-147, sheets I and 2.   '
The USAR evaluates pressure boundary failure, but the aspects of pressure boundary design are -
                                      ~
unchanged and remain qualified. There is no increase in the probability of a malfunction, since the  l ASME code qualification is maintained. Drywell leakage would be the consequences of a pressure boundary malfunction. The flow passages in the valves, piping and fittings that make up the Drywell penetrations are unchanged. Therefore, the change will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This change will have no affect on the probability of the accident, since this piping functions to lessen the consequences of a Loss-of-Coolant Accident. Changing the valves from closed to locked closed has no impact on leak tightness, therefore, this change will not increase the consequences of an accident previously evaluated in the USAR. The valves centain the             l accident rather than cause the accident. Since locking the valves closed will not change their function, they could not cause a new type of accident. The potential for mis-positioning is          ,
eliminated by locking the valves closed and the potential for failure of an active safety function   i valve is eliminated by manual control. Therefore, a malfunction different than previously            I evaluated in the USAR is inconceivable. The containment support function of the Drywell is           l maintained and there is no affect on the containment boundary. Therefore, this activity does not     I reduce a margin of safety as defined in the basis for any Technical Specification.
CLOSING OF BYPASS STEAM ADMISSION VALVE 1E51-F095 DURING RCIC SYSTEM OPERATION i
Document Evaluated: USAR Section 5.4 6                                          Log Number: 97-042 The purpose of the Reactor Core Isolation Cooling (RCIC) system bypass steam valve IE51-F095 is to prevent over-speeding of the RCIC turbine during RCIC start-up by limiting the steam flow to the turbine. Currently, the bypass steam valve IE51-F095 is closed from the Main Control Room (MCR) under no steam flow conditions after the turbine is tripped. This activity changes closing the valve lE51-F095 (under steam flow) prior to tripping the turbine. GE design specifications states that the bypass valve IE51-F095 is capable of opening and/or closing against full pressure.
The function of the bypass valve is not affected by closing it prior to tripping the turbine. When the bypass steam valve is closed, RCIC turbine steam admission valve IE51-F045 will be fully open. Therefore, the change will not increase the probability or the consequences of a malfunction previously discussed in the Updated Safety Analysis Report (USAR). The change will not increase the probability or the consequences previously evaluated in the US AR, because the functions of the RCIC system and the bypass valve do not change. Because the bypass valve is capable of closing under maximum differential pressure, the change will not create the possibility of a new accident or a malfunction of equipm:nt not evaluated previously in the USAR. The ir, tended function of the RCIC system is not affected; therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.

1 At'.achticnt 2
                                                                                                          ' l to U-603276 Page 22 of 252 RCIC INJECTION VALVE DISC MODIFICATION Document Evaluated: ECN 29598                                                   Log Number: 97-044 Engineering Change Notice (ECN) 29598 involves drilling a small hole in the Reactor Core Isolation Cooling (RCIC) Injection Shut Off Valve disc (1E51-F013) to resolve gate valve pressure locking concerns. Pressure locking occurs when high pressure water is able to leak past one of the valve seats into the valve bonnet while the valve is closed. This trapped fluid can increase the force necessary to open the valve beyond the capability of the actuator during unseating. The result could be the failure of a motor operated valve to perform an active safety function. Valve IE51-F013 is a normally closed valve with an automatic safety function to open in the event of a RCIC injection. Another automatic safety function of valve 1E51-F013 is to close (if open) to isolate RCIC injection and provide Containment Isolation. Adding a small hole would provide a relief path from the valve bonnet to the RCIC injection piping, preventing pressure locking.
With this activity, the scquence of events is unchanged, except that reduction in reactor pressure would not jeopardize the opening capability of the valve, actually decreasing the probability of a malfunction. The modification does not change the consequences or magnitude of any malfunction, because there is no change in the response of IE51-F013 to a failure as an active safety function. This activity will not increase the probability of an accident, because the failure of the RCIC Injection Valve to perform an active or passive safety function is not an initiating event for any accident described in the Updated Safety Analysis Roort (USAR). The'USAR evaluates several scenarios in which the failure of the RCIC system is assumed. The safe shutdown of the plant or the release of contamination in excess oflegal limits is not a possibility during the failure or unavailability of any single division of the RCIC system. Therefore, the modification of the RCIC Injection Shutoff Valve disc does not affect the consequences of any accident described in the USAR. This modification does not change the design function or operation of the RCIC system or uv ofits components or introduce any new components. The result of this modification is a greater assurance that valve 1E51-F013 will operate as described in the USAR.
The failure mode of the valve has not changed. Therefore, there is no new type of accident created as a result of this modification. This activity does not affect compliance with any Technical Specification. Therefore, this activity does not reduce a margin of safety for any Technical Specification.
ADD PIPING UNIONS TO OFFGAS REGENERATION SKID DRAIN TRAPS.
Document Evaluated: ECNs 30110 and 30116                                        Log Number: 97-047 Engineering Change Notices (ECNs) 30110 and 30116 added piping unions to the piping attached to the drain trap (1N66D006A/B) on the Offgas Regeneration skid (IN66Z004). These unions will make drain trap removal easier for maintenance. This modification will have no impact on regeneration skid functions or operation. The unions will be installed in accordance with the same installation standards used throughout the non-safety related hardware in the plant. Leak testing

1 0
Attachment 2 to U-603276 Page 23 of 252 will ensure proper installation. This system and the affected components are not associated with the assumed initiators of any Updated Safety Analysis Report (USAR) evaluated accident. Nor do these components function in a mitigatory capacity that would efRrt the resultant dose consequer.:es of any evaluated accident. Therefore, this activity will not increase the probability of a malfunction of equipmem previously evaluated in the US AR. Consequences of a malfunction of equipment important to safety previously evaluated in the USAR are not affected by installing piping unions on the skid. Potential leakage of offgas through the piping unions was discussed with the analyzed failure of the offgas piping pressure boundary. However, these components tneet the material and dr ~gn standards required for this system and their failure probability is no greater than that of the original system components. Therefore, the change will not increase the probability of an accident presiously evaluated in the USAR. Gaseous leakage through the unions or stmetural failure would not result in release of offgas greater than that already analyzed in the USAR. The only conceivable accident resulting from the addition of unions would be leakage of ofTgas from the piping. Since this was stready analyzed in the USAR, the change will not create the possibility of an accident or malfuncthn of equipment of a ditibrent type than any evaluated previously m the USAR. This activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
l MANUAL COMPENSATORY ACTION FOR DIVISION 1,2, & 3 DEGRADED VOLTAGE RELAYS Document Evaluated: OD l-97-01-276, Action #1                                    Log Number: 97-054 This Safety Evaluation was performed for manual Compensatoly Action number 1 established as
,            part of the Operability Determination (OD) for Condition Report (CR) 1-92-04-031.
]            CR l-92-04-031 identified that the voltage setpoints of the Division 1,2, and 3 degraded voltage relays were set too low and were not suflicient to ensure proper low voltage protection for all Class 1E equipment. As a result of that CR, Clinton Power Station (CPS) installed an annunciator alarm for the Divisions 1,2, and 3 4.16-kV buses and established operator actions in crder to ensure adequate voltage for proper operation of all Class IE equipment. CPS annunciator procedure 5008.05, "4kV Bus Low Voltage," states: "IF 4.16 kV Class lE bus voltage falls below 3890 VAC, THEN Start the Diesel Generators (DGs) and transfer the buses           a to the DGs.. "
           - The operator action does not affect equipment malfunctions that are described in the Updated E.tfety Analysis Report (USAR). Therefore, the operator action will not increase the probability or consequences of a malfunction of equipment important to safety evaluated previously in the USAR. The USAR does not evaluate an accident concurrent with degraded voltage. The method ofisolating from a degraded source is changed, but not the probability of needing to isolate.
Therefore, this change does not increase the probability or consequences of an accident previously
: e. valuated in the USAR. This change does not involve any manipulation of pe it equipment not previously discussed in the USAR. As a result, operation of this equipment can not initiate an accident not previously analyzed. However, due to this change, the operator has to respond for all three divisions simultaneously, if the operator does not respond properly or quickly enough,

p Attachment 2 to U-603276 Page 24 of 252
     . there is a possibility of creating a malfunction cf equipment important to safety. Therefore, this change is deemed to involve an unreviewed safety question. In accordance with 10 CFR 50.59, a license amendment has been submitted to the NRC seeking review and approval of the proposed change (reference Letter U-602794, dated July 20,1997). In addition, plant modifications are in progress to permanently resolve the degraded voltage issue (reference Modification AP-37 and
       'AP-3 8).'
y        CORRECTION TO PHYSICAL LOCATION OF AREA RADIATION MONITORS
      ' Activity Evaluated: US,AR Change 7-160                                                                                                                                                        Log Number: 97-092 R/l Updated Safety Analysis Report (USAR) Change Package 7-160 revised locations of various Area Radiation Monitors (ARMS) as described in the USAR. . Changes to the USAR included
     . clearlyidentifying actual areas monitored, changing the location of several ARMS to reflect actual locations .vithin the plant, revising figures to indicate correct locations, and removing a reference to ARMS within Residual Heat Removal (RHR) rooms "A" and "B" from USAR Section 6.2.4.2.
These changes only affect the monitor's location as depicted on USAR figures or better defines the areas monitored. The changes do not impact the ARM's ability to monitor the existing, defined area's . The ARMS are passive devices since they perform no control or trip function. No system response characteristics or design limits were affected by these changes. The ARMS do not provide information or protection for any safety related equipment and the failure modes of the ARMS remain the same. Therefore, this activity.does not increase the probability or consequences of an accident or malfunction of equipment imponant to safety previously evaluated
      . in the USAR.' Adverse interactions with other systems or operational transients were not isroduced by these changes. The reference to ARMS within RHR rooms "A" and "B" was
     ' deleted, because there are no ARMS located in either RHR Room "A" or "B" USAR sections
     . 5.2.5,7.6,1.4 and 7.6.2.4 states that ponions 'of the Emergency Core Cooling System outside                                                                                                                               ,
primary containment is provided primarily by floor and equipment drain sump leak detection                                                                                                                              j
     . system and by the equipment area high-temperature detection system,'neither of which are dependent upon ARMS to perform their design functions. Since the ARMS are not required for monitoring leak detection in the RHR rooms, their removal from the rooms does not create the possibility of an accident or malfunction of equipment important to safety of a different type than L        previously evaluated. Funber, this activity has not affected the failure probability of the RHR system within this room; nor -does removal of the monitors impact the potential consequences of
[    - RHR system leakage which would be diagnosed based on leakage detection system response.
;        None'of the detectors affected by this change are discussed in the Technical Specifications.                                                                                                                            l Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification
                          '4 1

w,
* Attachment 2 to U-6032' ~
Page 25 of 252 CYCLED CONDENSATE SYSTEM DRAWING DISCREPANCIES y Document Evaluated: M05-1012, Sheets 1 and 5                                    Log Number: 97-095 This change resolves discrepancies that exist between the configuration of valves in the Cycled
            ' Condensate (CY) system and drawing M05-1012. These discrepancies include mislabeling, lack                -
oflabels,'and different valve positioning. Updated Safety Analysis Report (USAR) Figure 9.2-11 corresponds to drawing M05-1012.
The condensate storage facilities provide water to nuclear safety-related systems, but do not perform safety-related functions themselves. The discrepancies may affect certain 'tagouts or valve line-ups, but they will not reduce the performance of containment isolation valves.-
Therefore, the probability or consequences of a malfunction of equipment important to safety are unchanged. Chapters 6 and 15 of the USAR discusses possible accident scenarios; the changes being made.do not affect these scenarios. Furthermore, the containment isolation system is not affected by these discrepancies. Therefore, these drawing changes do not increase the probability or consequences ot'an accident previously evaluated in the USAR. Because these discrepancies do not reduce the performance of the valves as system boundaries, there is not a potential for creating an accident or malfunction of equipment important to safety of a different type than previously' evaluated. These changes do not reduce a margir, af safety as defmed in any Technical
             - Specifications, because the fuel cladding, reactor pressure boundary, and containment fission -
product barriers are unchanged.
RESERVE AUXILIARY TRANSFORMER PRIMARY TAP CHANGE Activity Evaluated: ECN 30102                                                  Log Number: 97-102 Engineering Change Notice (ECN) 30102 repositions the primary winding tap for the Reserve ..
Auxiliary Transformer (RAT). Repositioning the primary winding tap will decrease the turns ratio, will result in an increase in transformer secondary voltage, which will result in a higher plant . i
             - AC Distribution system voltage. This activity will help ensure that adequate system voltage is supplied to all Class IE equipment. The increase in the RAT secondary voltage will result in the following impactk to the facility as described in the USAR: change the worst case transient / steady state voltage outlined in Updated Safety Analysis Report (USAR) Section 8.2.2.1; increase the fault currents outlined in USAR Tables 8.3-19 and 8.3-20; require balancing of the 4.16KV loads             !
and thus, change the description of plant operation as describe ! in USAR Sections 8.1.3.1 and              !
      <       8.3.1.1; decrease the plant loads that tap off the Main Power Transformers (MPT), thus                      j increasing the maximum loading of the MPTs as described in USA R Sections 8.3.1.1.1; and j
designating the Emergency Reserve ' Auxiliary Transformer (ERAT) as the preferred power supply              j to the Class IE busses when the plant is shutdown under lightly loaded transmission conditions,             i thus changing the descriptbn of plant operation in USAR Section 8.3.1.1.2.                                  !
l The generator load rejection accident is evaluated in USAR Section 15.2.2. A loss of any Main Power Transformer (MPT) is an assumed initiating event for this accident. ECN 30145 Nstalled

4 Attachment 2 to U-603276.
Page 26 of 252
    ' MPT monitoring instrumentation to ensure that an overload condition does nd scur at the MPT.
Loss of AC Power accident is evaluated in USAR Section 15.2.6, the loss of a Unit Auxiliary Power (UAT) . transformer, either by equipment failure or actuation of the transformer protective
    ; circuitry, is considered an initiating event for this accident. The RAT is designed to provide rated power at all the tap settings, and would not cause the transformer protective circuitry to actuate.
The UATs and MPTs are still operated within their design ratings. The change in the RAT primary taps alone results in an increase in voltage to Class IE equipment. Calculation 19-AK-06
. determined that with implementation of the plant modifications listed above, the worst case fault currents and voltages at the Class IE loads would not exceed the design ratings of the Class IE equipment.. Since all Class IE components are operated within their design ratings, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This modification provides protection against degraded voltage conditions while ensuring that Class IE loads do not exceed their design l parameters.. This modification does not introduce any new failure mechanisms to the Class IE equipment it supplies Therefore, his modification does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated.
The Technical Specification Bases discuss frequency, voltage and redundancy specifications for the AC distribution system. Since this m6dification adheres to these specifications, it does not reduce a margin of safety as defined in the basis for any Technical Specification.
DRY SOLID WASTE PACKAGING AND STORAGE ACTIVITIES AND EQUIPMENT
    '. Activity Evaluated:: USAR Changes 7-269 and 8-317                                  Log Number: 97-113 R/l The use of off-site processing has become the standard inode of operation for the dry active waste processing. This change to the Updated Safety Analysis Report (USAR) addresses the dry active waste sorting operations at the soning enclosure, the process to package waste materials for off-site shipment from the facility, and temporary storage of dry active waste outdoora, but within
     ' the protected area. Revision I to this Safety 9 valuation addresses the failure to include USAR Figu'r e 1 L4-1 in the original evalua't ion, the potential impact on the Main Puwer Transformers,-
need to implement some form of control over the maximum number of containers assumed in aralysis, and inappropriate use ofin-progress analyses. Thirteen containers are the maximum number ofloaded containers that will be stored at any time. This maximum number is controlled by Clinton Power Station Procedure 7013.20. The area is serviced with both oil separator and storm water drains to prevent accumulation of water from either rain or Main Power Transformer (MPT) deluge system actuation. The storage location is at a higher grade than the drains,
     . assuring water does not accumulate in the area. Safe operating distances are maintained in the vicinity of the MPT, Reserve Auxiliary Transformer (RAT), and Emergency Reserve Auxiliary Transformer (ERAT),' providing assurance that the container does not contact the MPTs or the pylons of the RAT during conduct of routine operations or high wind and tornadic conditions.
The handling and storage of radioactive waste materials is located such that it can not affect
     - overall ' system performance, system response characteristics, system design limits, system operational transients, or system interactions in any_ manner that would increase the probability or consequences'of an ' accident or malfunction of equioment important to safety previously evaluated

y
    .r, 4                                                                                           Attachment 2 to U-603276 l                                                                                             Page 27 of 252
            .in the USAR. The possible accident scenarios are bounded by existing USAR design basis events and no new pSnt equipment was introduced by this change. Therefore, this activity does not f             create the possibility of an accident or malfunction of equipment important to safety of a different L             type than previously evaluated. The dry active waste management system is not directly
           . represented. in the Technical Specifications.: Limitations on the number of containers (hence l           : aggregate source term) ensure that the Radiological Limits of Technical Specifications 5.5 4.g and      -
L           l 5.5.4.I are not impactedi Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
CONDENSER PIT NOT FLOOD PROOF UP TO ELEVATION 715' AS DESCRIBED IN         .
            . THE USAR Document Evaluated: CR 1-96-12-150                                     ' Log Number: 97-122 This Safety Evaluation accepts'the "as-is" configuration of the condenser pit located in the turbine building basement; thus, resolving the Updated Safety Analysis Report (USAR) discrepancy identified in Condition Report (CR) 1- % 150 The USAR staies in several locations that the condenser' pit is capable of contain'mg floods within it up to'elevatian 715' Actually, two drain
            , lines provide a flow path from the condenser pit for floods above approximately 709'-5". Section D3.6.4 of the USAR evaluates the maximum flood for the condenser pit for a circulating water expansion joint faihire. This analysis indicates that the entire turbine building basement would flood up to an elevation of 715'
             -Safety related equipment required to ' safely shutdown the plant are not affected by flooding events
            ' in the basement of the turbine building. The condenser pit drain lines create the possibility of flooding the turbine building basement without first causing the condenser pit level to exceed the point at which level switches trip the circulating water pumps. These events involve break flow rates much smaller than the circulating water expansion joint failure discussed in the USAR. It would take much longer to fill the basement of the turbine building at the lower flow rates. The area, depth, and total volume of the turbine bt.ilding flooded would be the same regardless of the
             . presence of the drain line leakage paths from the condenser pit. Therefore, this change does not
            ' create the possibility of a new type of accident or equipment malfunction from what was previously evaluated. Circulating water line breaks and turbine building floods are not addressed
            ' in the Technical Specifications,'therefore, this activity does not reduce a margin of safety as
     .      . defined in the basis for any Technical Specification.
4 RHR SYSTEM DRAWING DISCREPANCY Document Evaluated: CR l-%-11-034                                               Lg Number: 97-125 During the review / investigation process of Condition Report (CR) 1-96-11-034 regarding tightening of bolts on an Integrated Leak Rate Test boundary flange, plant drawing discrepancies were identified.L After initial startup, strainers were removed from system piping and spacer plates were installed in their place. Later, the spacer plate positions were interchanged with flow orifice's r

E y                                                                                              Attachment 2 to 11-603276 Page 28 of 252
           . to correct pipe vibration problems. Field Engineering Change Notices (FECNs) 9332,10117 and
           ' 12240 were issued to revise the associated drawings; however, the FECNs did not show the correct location of the spacer plates. The following drawings / figures were changed to show the
           ' correct 1.ccations of the' spacer plates: _ Updated Safety Analysis Report (USAR) Figure 5.4-13 sheets 1,2, and.3; USAR Figure 3.6-1 sheets 62,63, and 64; USAR Figure 6.2-147 sheets 1,2, and 3; Piping and Instrumentation Diagram Residual Heat Removal (RHR) System M05-1075
              ' sheets 1,2, and 3; Operational Schematic RHR System 05-1075 sheets 1 and 2i RHR Piping M06-1075 sheets 3, 8,' 15,19, and 25.
           ; The spacer plate location has .no impact on the safety function of the RHR system. Since, no
_ physical change is being performed to any component, this change did not increase the probability or consequences'of an accident or malfunction of equipment important to safety evaluated previously in the USAR. All saDty functions of the system will continue to be met; therefore, this -~
change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The RHR System will continue
           . to meet all requirements and provide all specified functions assumed in the Technical
              . Specifications; therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
ERAT-TAP SETTING CHANGE FROM TAP #3 TO TAP #2 Document Evaluated: ECN 30169 '                                                 Log Number: 97-126 Ehgince-ing Change Notice (ECN) 30169 repositions the primary winding tap for the Emergency
    ,         Reserve Auxiliary Transformer (ERAT)n This activity will help ensure that adequate system voltages are supplied to all Class IE equipment. This change is similar to the change made for the
           ' Reserve 'Auidiary Transformer (RAT) reported in Safety Evaluation 97 102. Loss of AC Power accident is evaluated in Updated Safety Analysis Report (USAR) Section 15.2.6, the loss of a Unit AuxiliaryTransformer (UAT), either by equipment failure or actuation of the transformer L          - protective circuitry, is considered an initiating event for this accident. The ERAT is designed to provide rated power at all the tap settings, and would not cause the transformer protective circuitry to actuate. The UATs and Main Power Transformers (MPTs) are still operated within their design ratingsJ The change in the ERAT secondary tap settings alone results in a decrease in                3 voltage t , Class IE~ equipment; however, the voltage supplied is well withm the required range                 1 l cnd the protective relaying is unaffected by this change. Station calculations have further determined that when the ERAT transformer taps are changed, Class lE components would not see an overvoltage condition. Therefore, since all Class IE components, are operated within their              ,
design ratings, this activity does not increase the probability or consequences of an accident or I
malfunction of equipment important to safety previously evaluated in the USAR. This modification provides protection against degraded voltage conditions while ensuring that all Class IE loads are not operated outside of their design parameters. This modification does not introduce any new failure mechanisna to the Class IE equipment it supplies. Therefore, this activity'does not create' the possibility an accident or malfunction of equipment important to safety of a different type than previously evaluated. The Technical Specification Bases discuss

       '.                                                                                        Attachment 2 to U-603276 Page 29 of 252 0 :.
frequency, voltage and redundancy specifications for the AC distribution system. Since this modification adheres to these specifications, it does not reduce a margin of safety as defined in the basis for any Technical Specification.
               - REPLACEMENT OF REGULATING TRANSFORMER WITH NON-REGULATING TRANSFORMER AND TRANSFORMER TAP Document Evaluated: Modification AP-032 Log Number: 97-133 Modificatior. AP-028, Safety Evaluation (SE) Log Number 96-015, installed new undervoltage relays and new regulating transformers to address degraded voltap concerns. AP-028 also had additional' changes that were also evaluated in SE 96-015.H Temporary Modification 97-017, SE Log Number 97-057, re-connected the non-regulating transformers, which AP-028 had abandoned in place, and in turn abandoned the regulating transformers in place. This modification, AP-032, makes permanent Temporary Modificatior. 97-017. This modification also makes changes to the tap settings in the distribution transformers inside Motor Control Centers (MCCs) 0AP55EB, OAP57E, and 1 AP75E, In addition, internal to the MCCs, this activity switched certain individual loads from non-regulated distribution panels to regulated distribution panels.'
Updated Safety Analysis Report (USAR) Section 15.2.6 evaluates Loss of Offsite Power. The
                ' degraded voltage setpoints were selected to be low enough to prevent inadvertent power supply tran'sfer, yet high enough to ensure that sufficient power is available to the required equipment.
               . This modification brings Division 2 into compliance with General Design Criteria (GDC) 17 for electrical power sources by providing minhnum required voltage to loads down to the'120 volt level while at the dropout setting of the degraded voltage relay. This change ensures the plant response is as described in Chapters 6 and 15. The changed loads retained the divisional and physical separation of channels described in US AR Sections 7.6.1.2.5.1 and 11.5,2.1.6. Illinois Power (IP) calculation 19-AI-62 verified proper breaker coordination and the supply breaker to the regulated distribution panel was verified to be adequate for the additionalload. The switched loads were not affected by the switch from non-regulated to regulated power supplies. Although, the changes to the tap settings of the distribution transformers were not included in either AP-028
               - or Temporary Modification 97-017, the analyses of SE Log Numbers 97-057 and 96-015 remain
               . valid for the re-connection of the non-regulating tram. formers and abandonment of the regulating transformers. The revised tap settings for the distribution transformers ensure that that minimum required voltage is provided to all loads. IP. calculation 19-AJ-72 determined the new tap settings and works in conjunction with the changes made to the Reserve Auxiliary Trancformer (RAT) and
:                . Emergency Reserve Auxiliary Transformer (ERAT). The changes to the RAT and ERAT were Li            accomplished with ECNs 30102 and 30169 (SEs 97-102 and 97-126 respectively). Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
By bringing the plant into compliance with GDC 17, a degraded voltage condition will be
                ' automatically mitigated by the second level undervoltage relays. The tap setting changes will ensure minimum required voltage is supplied to allloads. This modification maintains divisional 4
L..

                                                                                              ' Attachment 2 to U-603276
                                  '                                                             Page 30 of 252 e           iseparation and seismic qualification. Therefore, this activity does not create the possibility of an
              . accident or malfunction of equipment important to safety of a different type than previously evaluated in the USARETechnical_ Specification Bases 3.8.9 and 3.8.10 require distribution
              . systems to provide the capacity, reliability, and redundancy to support Engineered Safety Feature systems, and other plant systems. This modification along with the others referenced in this .
Safety Evaluation restore the ability of the distribution system to meet those requirements.              -
              - Therefore, this modification does not reduce a margin of safety as defined in the basis for any Technical Specification.
              - CIRCULATING WATER SYSTEM SCALE INHIBITOR TUBING REPLACEMENT Docurpent Evaluated: Temporary Modification 97-036'                             Log Numberf 97-136 Temporary Modification (TM) 97-036 will remove the 0.375 inch-18 BWG Stlainless Steel (SS)
              - tubing, needle valves (ICWO31 A & C) and connectors at valves ICWO32A and ICWO32C, then
              ; install 0.5 inch-Sch 80, A106-Gr.B,2 inch long pipe nipples with pipe caps at these valves. The
              . tubing was used as a quill and passed through valve ICWO32A(C) into the discharge piping of Circulatirig Water (CW) Pump ICW0lPA. These flow paths (ICW44AB and CB) will be restored to original CW system configuration before using these pipes for scale inhibitor injection points. This replacement will eliminate the risk of tubing failure that could require CW pump shutdowns, until repaired. To shutdown a CW pump, a reduction in station power is required which has the potential to introduce a plant transient.
The 0.5. inch-Sch 80, A106-Gr.B. 2 inch lo_ng pipe nipples and pipe caps are as specified in Pipe Design Table (PDT) 002. The pressure and temperature rating for PDT 002 components exceed the CW. system conditions.' Section 10.4.5.3 of the Updated Safety Ana ysis Report (USAR),
evaluates a circulating water pipe expansionjoint rupture at the inlet or outlet of the condenser water boxe's. The installation of the piping does not affect this accident. Therefore, this TM will
               - not increase the probability or consequences of an accident or malfunction of equipment important to safety evaluated previously in the USAR. Hypotheticalleakage from a piping failure in a 0.5 inch diameter line is estimated to be less than twenty gallons per minute. This leakage would be discharged to the CW Valve Pit; The CW Valve Pit Sump Pumps each have a capacity of eighty gallons per minute, therefore, flooding would not occur. This change will not create the
              ; possibility of an accident or malfunction of equipment of a different type than previously evaluated in the USAR. Even considering a hypothetical failure of the piping, there is no affect on the approved acceptance limits of the licensing basis. Therefore, this temporary modification does not reduce a marg n of safety as defined in the basis for any Technical Specification.
g]
1 4
i-y-
d.

Attachment 2 to U-603276 Page 31 of 252 ALTERNATE POWER TO THE TEMPERATURE CONTROLLER RACK IN PANEL 1H13-P867 Document Evaluated: Temporary Modification 97-037                                Log Number: 97-139 Temporary Modification (TM) 97-037 addresses one of the overvoltage concerns raised during the review of the circuits fed from the 120V distribution panels which are having their supply transformer taps changed to provide a 5% increase in voltage. The power supply for the temperature controller rack in the electronic control module portion of the H202 monitoring system has limitations on voltage input. The power supply provides operating voltage to the temperature controller rack which controls the heat tracing for various sampling lines in the system. The TM uses the control circuit of the air complessor for the H202 sampling panel (ICM01SA) to energize the power supply for the temperature controller rack in panel IH13-P867. This maintains the divisional power requirements of the equipment as well as the description of the containment monitoring system in the Updated Safety Analysis Report (USAR).
However, schematic diagrams, listed in USAR Table 1.7-1, for the system are impacted by the revised circuit wiring.
This temporary modification provides a Class lE power source to this power supply that will not subject it to over voltage. The control circuit has been evaluated to ensure that, in addition to providing the required power to power supply, the normal function will not be impaded by this TM. The H202 monitoring equipment is used post-accident to identify hydrogen concentration in the Containment and Drywell. This TM prevents one failure mode for the system that resulted from the tap change of the 120V distribution transformer. However, Chapter 15 of the USAR does i at identify any accident scenarios that are initiated from failure of these panels. Therefore, this temporary modification does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This TM is contained within existing circuits of the H202 monitoring system and maintains the syst6m functionality. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This activity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specificationc TEMPORARY POWER TO TEMPERATURE CONTROLLER RACK IN PANEL IH13-P868 Document Evaluated: Temporary Modification 97-038                                L,og Number: 97-140 Temporary Modification (TM) 97-038 addresses an overvoltage concern raised during the review of the circuits fed from the 120V distribution panels. This review was prompted in a ponse to a planned increase in the load tap setting on the 4kV winding of the Reserve Auxiliary Transfarmer.
The 120V power supply for the temperature controller rack in the electronic control module portion of the Hydrogen-Oxygen (H202) monitoring system has limitations on voltage input.
The power supply provides operating voltage to the temperature controller rack which controls

I Attachment 2 to U-603276 Page 32 of 252 the heat tracing for various sampling lines in the system. The TM uses the control circuit of the air compressor for the H202 sampling panel (1CM01SB) to energize the power supply for the temperature controller rack in panel IH13-P868. This maintains the divisional power requirements of the equipment as well as the description of the containment atmosphere monitoring system in the Updated Safety Analysis Report (USAR). However, schematic diagrams, listed in USAR Table 1.7-1, for the system are impacted by the revised circuit wiring.
The H202 monitoring equipment is used post-accident to identify hydrogen concentration in the Containment and Drywell atmosphere. This TM will prevent one failure mode for the system that resulted from the tap change of the 120V distribution transformer. There are no accident scenarios that are initiated from failure of these panels or supplied equipment. The control circuit has been evaluated to ensure that, in addition to providing the required power to power supply, the normal function will not be impacted by this TM. Since this evaluation determined that both functions could be performed within the ability of the unit, the TM would not change design basis equipment malfunctions. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
This TM i; contained within existing circuits of the H202 monitoring system and does not impact any equipment outside of this system. Also, this activity maintains the existing system functionality. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. This activity does not degrade ccmpliance with any Technical Specification safety limit, limiting safety system setting, or limiting ondition for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REMOVAL OF POWER FROM RHR 'B' SHUTDOWN COOLING SUCTION VALVE WHEN NOT NEEDED FOR PLANT OPERATIONS Document Evaluated: ECN 30211                                                  Log Number: 97-143 Engineering Change Notice (ECN) 30211 removes power from the 'B' Residual Heat Removal (RHR) System shutdown cooling suction valve (IE12-F006B) when not needed for shutdown
    - cooling operations. This change will prevent spurious uncontrolled operatien of the valve, which could adversely affect plant shutdown capabilities. Also, this change adds an alarm bypass switch, to prevent a nuisance alarm in the control room when the valve is cLsed and power is removed.
The purpose of the change is to ensure valve 1E12-F006B is in the closed position during modes of operation that do not require the valve to be operated.
Valve IE12-F006B interlocks with another valve. A hot short caused by a fire could allow IE12-F006B to open. If IE12-F006B is open, the interlock would not allow the other valve to be opened from the control room. By removing power to valve IE12-F006B, during modes of
'    operation that do not require operation of the valve, the probability of a malfunction of equipment important to safety under a design basis fire will decrease. 'The change requires some operator actions outside the Main Control Room f o get into normal shutdown cooling. However, the Shutdown Cooling Outboard Suction Isolation Valve requires the same operator actions as valve

       ,-.                                                                                         Attachment 2
                                                                          <                        to U-603276 Page 33 of 252
                 <1E12-F006B to establish this mode of shutdown cooling. These actions are proceduralized,
                ; considered no'n-challenging, and do not have time constraints that may contribute to an operator error 1 Removal of power to the valve has no affect on the probability or consequences of an
: accident; it only affects the method of coping with a de_ sign basis fire. Basic operation of the Residual Heat Removal shutdown cooling system is unchanged. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.                  -
aUSAR," APPENDIX F, TABLE FA-1 DISCREPANCY RESOLUTION -
s
: a.      ,
                 ' Document Evaluated: USAR Appendix F, Table FA                               Log Number: 97-144
                   'A discrepancy was found in Note 12 of the Updated Safety Analysis Repon (USAR),
1 Appendix F, Table FA-1. The information was based on a non-conservative interpretation of
                '_ Nuclear Regulatory Commission (NRC) guidance contained in Question 5.3.10 of Generic Letter 86-10. The guidance provided by the NRC applies to alternate and dedicated shutdown
: capability. However, Clinton Power Station applied this guidance to non-alternate shutdown fire areas. ' As a result, an evaluation for electrically operated valves whose resolution was based on
                  ~ Note 12 was' performed. The following valves have a revised resolution: 1S E 11A/B,-
ICC071/72, ICC073/74, IE12-F094, IE12-F096, IE12-F%0A/B, IE12-F075A/B, IE12-F040/049, IG33F028/34/35/46, and IE51-F022/59J This change re-evaluates the Safe Shutdown Analysis and' demonstrates conformance to the Title 10, Part 50 of the Code of Federal Regulations, Appendix R requirements pertainitig to hot shorts. The evaluation assures compliance to' Appendix R requirements, and thereby, this change does not increase or create the
                 . probability or consequences of an accident or a malfunction of equipment important to safety
?
: previously evaluated in the USAR.- ' Demonstrating conformance to the 10CFR50, Appendix R requirements assures that safe shutdown under the design basis fire can be achieved. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REVISION TO USAR FIGURE 8.3-7 TG REFLECT AS-BUILT CONDITIONS Document Evaluated: USAR Change 7-279                                         ' Log Number: 97-147 l
                        .                               .                                                                i
                  ' This change supports revisions to Updated Safety Analysis Report (USAR) Figure 8.3-7, previously made to the base drawing, E02-lDC06. The base drawing and the as-built field condition show a fused connection with the battery and battery charger to 125VDC Motor
                  - Control Center (MCC) ID, whereas USAR Figure 8.3-7 shows a circuit breaker connection. The           ,
base drawing for High Pressure Core Spray Division 3 battery shows breaker CB-6 and ammeter          1 M17 in a reversed configuration from that shown on USAR Figuie 8.3-7.
                  ' USAR'Section 8.3.2.1.1 states that only Division 1, Division 2, and Division 3125-VDC
                  ' subsystems are regnired to be considered for safe shutdown analysis of the plant. Given that the Division 3 change is editorial only, the proposed change will not increase the probability or consequences of equipment important to safety evaluated previously in the USAR. Accidents

                                                                                         ' Attaciunent a
                                                                                        . to 'U-603276
                                                                                        ; Page 34 of 252 associated with Division'4 of the 125-VDC system are not analyzed in the USAR. 'Therefore, this change will not increase the probability or consequences of an accident previously evalated in the
  < : USAR.E The first change enhances system performance by correcting a problem of brerier coordination which existed with the original design and the second change is an editorial change depicting the ammeter connection point. ' Neither change creates an accident initiating event;
: therefore, the changes will not create the possibility or consequences of an accident of a different         -
type than any evaluated previously in the USAR. A review of the Technical Specification Bases
      -indicates that the changes have no_ affect on the information in Bases 3.8.4,3.8.5, and 3.8.6.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical
: Specification.
NON-LICENSED OPERATOR' TRAINING PROGRAM DESCRIPTION -
Document Evaluated: Non-Licensed Operator Training                                  Log Number: 97-148 Program Description -
Except for refueling activities, specific task certification of non-operations ~ personnel is not allowed under the Non-Licensed Operator Training Program Description (NLO TPD) revision 10.
Revision 11 provides for outage augmented task c'ertification of non-operationr personnel to perform two specific tasks normally performed by Non-Licensed Operators Ged through C
     . Area only. Revision 11 adds the following tasks to the NLO TPD: " Vent n Control Rod Drive Hydraulic Control Unit During a Plant Shutdown" and "Precharge a Depressurized Rod Drive Hydraulic Control Unit Accumulator During a Plant Shutdown Condition." The two tasks will be
     ' performed in accordance with Clinton Power Station (CPS) Procedure 3304.01, " Control Rod Hydraulic and Control." Classroom instruction and On the Job Training and Evaluation (OJT/E)
     - are requirements specified in revision 11 of the NLO TPD. The Control Rod Drive Hydraulic
     ' Control Unit (CRD HCU) Valves are restored to proper configuration when the applicable                         {
procedure sections are completed. CPS Procedure 3001.01, " Approach to Critical," requires peiformance of CPS 3001.0lV001. CRD HCU valves are returned to their proper configuration prior to startup per proper execution of CPS 3304.01, CPS 3001.01, and CPS 3001.0lV001.
Proper execution of these procedures ensures that performance of these CRD HCU related tasks ha e no impact on:L equipment malfunctions or accidents evaluated previously in the Updated
     - Safety Analysis Report (USAR), creating a new type of accident or equipment malfunction, or the margin'ofsafety.
CONFORMANCE WITH GDC 17 - ELECTRIC POWER SYSTEMS Document Evaluated: Tech Spec Bases Change -                                        Log Number: 97-151 The purpose of this change is to revise the basis for how Clinton Power Station (CPS) complies with General Design Criterion (GDC) 17. A section of GDC 17 requires that each of the two independent offsite power sources shall be designed to be available in sufficient time following a loss of all onsite AC power supplies and the other offsite electric power circuit, to assure that              3 specified acceptable fuel design limits and design conditions of the reactor coolant pressure                  j l

r:
r
       .                                                                                            Attachment 2 to U-603276 Page 35 of 252
: boundary are not exceeded. In the event of a loss of onsite AC electrical power, both the 138KV and 345KV offsite sources are considered to be available immediately. During certain grid-conditions, it has been predicted.that the voltage of either offsite source may be too low to be an l
immediate source. However, evaluations indicate that there is' sufficient recovery time to ensure
                 ; that fuel design limits and design conditions of the reactor coolant pressure boundary are not l-exceeded                           '                                                                 -
: l.                            .          .
<                EThe Updated Safety Analysis Report (USAR) does not e' valuate the consequences of accidents or
: equipment malfunctions concurrent with having only one of two offsite sources available.
Therefore, the probability.of an accident or a malfunction of equipment important to safety may L                    be slightly increased, but operator actions specified in the bases change will help ensure that
                 " GDC 17;will be met. Thus, these actions will help offset any potential increased probability or L consequences.? No new type of accident has been created, because the loss of offsite power is the bounding analyzed event. - No possibility of an equipment malfunction important to safety has E                    been created that has not been previously evaluated, because no new failure modes have been
                  - created.4 It has been interpreted that the Nuclear Regulatory Commission's (NRC's) " acceptance
                  . limit" is "two immediate sources," therefore, this change is a deviation from what was previously
                  ~ " accepted." This change is a reduction in the margin of safety as defined in the basis for the Technical Specifications.
t
: In accordance with 10 CFR 50.59, a license amendment has been' submitted to the NRC seeking review and approval of the proposed change (reference Letter U-602794, dated July 20,1997).
In addition, plant modifications are in progress to permanently resolve the degraded voltage issue
:(reference Modification AP-37 and AP-38).
                   . ' REVIEW STATUS OF M-053 AND ASSOCIATED CONDITION REPORTS Document Evaluated: CR l-97-05-189 -                                           Log Number: '97-153 During the 1988-1989 ' time frame, Modification M-053, a liquid Sodium Hypochlorite (NaOCl)
[                  isystem was installed. It was decided to not complete M-053 after problems were experienced-with the gravity transfer of NaOCl. Over time, other temporary modifications and plant --
modifications were installed to chemically treat the Circulating Water, Plant Service Water, j
l Potable Water, Shutdown Service Water (SX), and Demineralized Water Make-Up systems.
             .      ' Since the M-053 Modification Package has not been completed, various Updated Safety Analysis Rep 6rt (USAR) discrepancies have been ' discovered. - The SX system is the only safety related E                    ' system'which could be impacted by the status of the M-053 modification in its present condition.
p                     There are no electrical or meo.nical connections that have direct bearing on the SX system.
                    . .Along with a disabled annunuator tile in the Main Control Room, red line/ green line drawings make the operators aware of the modification status. The completion status of this system und its present configuration is unrelated to the' assumed initiators of any evaluated accident. This
: system's configuration and status do not have a deleterious effect on the SX system piping
                    . integrity, and _thus,' does not increase the probaU!ity of an evaluated moderate energy line break.
                   ; Nor does the status and configuration of this modification have a detrimental impact on

Attachment 2 to U403276 I
4 Page 36 of 252 equipment cidssified as important to safetyJAny previously evaluated accidents or malfunctions of equipment important to safety _will propagate in the same manner without influence by the' modification configuration. This modification configuration in no way degrades the ability of the SX system to_ perform its assumed Technical Specification function. Therefore, the present
        . condition of Modification M-053 does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the USAR, create the possibility of an accident or malfunction of equipment important to safety of a different type, or reduce a margin of safety as defined in the basis for any Technical Specification.
         . OFFSITE DOSE CALCULATION MANUAL CHANGE Document Evaluated: Offsite Dose Calculation Manual -                           Log Number: L 97-154 -
This change to the Offsite Dose Calculation Manual (ODCM) incorporates technical changes and corrects several obvious typographical errors. The editorial changes do not change the intent of
         . the ODCM and are consistent with regulatory requirements and Technical Specifications. The .
        - levels of radioactive effluent control are maintained and the changes do not adversely impact the
                                                       ~
           ' reliability of efIluent, dose, or setpoint calculations. The technical changes incorporated in this revision align the~ ODCM with Technical Specification 5.5.4.b liquid release limits, eliminate unnecessary sampling for the CCW when shutdown service water is not supplying cooling water
         = to the fuel pool heat exchangers, add CCW Process Radiation Monitors (PRMs) to the monitoring program, specify that the communication links are included in PRM channel check-
          ! operability, and add alternate flow rate measuring devices for the Standby Gas Treatment System
         ! gaseous effluent monitoring insvuments. These technical changes are consistent with regulatory              j requirements and Technical Specifications.                                                                 j
                 ~
: The technical changes incorporated bp this revision are related to surveillance and monitoring of
        ' radioactive effluents and calculations of dose to the public not to operation of plant equipment.
The probability or consequences of a malfunction of equipment important to safety are not affected by the' editorial and technical changes incorporated by this revision to the ODCM. The
        - ODCM provides the monitoring and calculation requirements to assess and report dose to unrestricted areas and the general public as a result of plant operation. The ODCM is not used to operate equipment necessary to mitigate accidents. There are no accidents associated with
        ; calculations for radioactive effluent dose evaluated in the Updated Safety Analysis Report
        - (USAR)J Therefore, these changes will not increase the probability or consequences of an accident previously evaluated in the USAR. Changes incorporated by this revision are being
                                    ~
made to ensure that dose resulting from radioactive effluents is accurately determined. Therefore, i this change does not create the' possibility of an accident or malfunction of equipment important to safety of a different type. This activity does not reduce a margin of safety as defined in the
        - basis for any Technical Specifistion, because these changes bring the ODCM into full compliance with the requirements of the Technical Specifications for reporting dose to unrestricted areas as a
        - result ofliquid and gaseous effluents.~                                                                      ;
e
                                                                        =                                              !

c Attachment 2 to U-603276 Page 37 of 252 REVISION TO USAR FIGURE 9.2-12 TO CORRECT NORMAL VALVE POSITION Document Evaluated: USAR Change 7-287                                              Log Number: 97-155       '
This change revises Updated Safety Analysis Report (US AR) Figure 9.2-12 to indicate that
:  Electrohydraulic Control (EHC) Cooler 1B cooling water outlet valve, IWT032B, is normally                 -
closed.' The vendor manual states that only one EHC cooler should be used while the unit is in l          -service. However, the Piping and Instrumentation Drawing M05-1057 Sheet 1 (associated with USAR Figure 9.2-12) shows both cooling water outlet valves as normally open. By changing the drawings to depict the IB cooler outlet valve as normally closed, the drawings, operating procedures and vendor manual will be in agreement.
Since there is no flow through the oil side of the out-of-service cooler, the change to USAR Figure 9.2-12 will have no efTect on the EHC oil temperature (which is maintained by a temperature control valve). Also, positioning the cooling water outlet valve in the closed position ensures proper temperature control of the EHC fluid and provides additional flow capa;ity to other components cooled by the Turbine Building Component Cooling Water (TBCCW) system.
Therefore, the probability or consequences of a malfunction of equipment important to safety evaluated previously in the USAR doe = not increase. The proposed change to a USAR Figure indicating a change in the valve position is in accordance with vendor requirements and has no effect on the ability of the EHC or TBCCW systems to function as designed. Thus, the proposed
           . change will not increase the probability or consequences of an accident previously evaluated in the USAR. The change does not adversely affect any equipment important to safety and will not I
create the possibility of an accident or a malfunction of equipment important to safety of a
;             different type than previously es atuated in the USAR. The drawing depicting the position of the valve is not addressed in the Technical Specifications or the associated Bases; therefore, this           3 activity does not reduce a margin of safety as defined in the basis for any Technical Specification.      ]
REACTOR RECIRCULATION PUMP SEAL LEAKOFF LINE CHANGES Document Evaluated: ECNs 30237 and 30238,                                        Log Number: 97-158 USAR Change 7-295 Engineering Change Notifications (ECNs) 30237 and 30238 change the Reactor Recirculation (RR) pump seal leakoffline to stainless steel l substitute ball valves for the existing globe valves, remove extra flanges and add a small slope in the line to assist drainage. These changes would help prevent the possibility of corrosion products migrating from the former pipe to the RR seal and potentially degrading its integrity. The valves were changed to ball valves for maintenance
           - purposes and availability of stainless steel. Removal of the flanges eliminates a potentialjoint failure. The intent of this change is to reduce the probability of a RR pump seal failure, there is no impact on the leakage measurement. The replacement components meet the material and design requirements for this system. Further, they have a demonstrated performance capability exceeding the existing components. Therefore, the change will not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the e
1 o                                                                                                                         !

Attaclunent 3 to U-603276 Page 38 of 252 Uhdated Safety Analysis Report (USAR). If the leakoffline should fail, the leakage would be
: collected by the drywell floor drain sumps rather than the equipment drains. If the seal fails, the
    ' material of the piping is irrelevant to the plant response to the event, Therefore, this activity does
    - not increase the probability or consequences of an accident or malfunction of equipment
     . important to safety previously evaluated in the USAR. There is' no plausible accident scenario that'could be created due to the proposed changes, because the piping is in the same configuration          .
except for4he material and slope. . Since the change directs pump seal leakage to the measuring device injust the same manner as before, no new possibility of a malfunction of equipment important to safety exists. Due to the fact that the seal leakage collection and the identified leakage allowable values are not changed, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
PLANT SERVICE WATER SYSTEM MINOR PIPING CHANGES Activitp Evaluated: ECNs 28161,28788,28789,28790,                             Log Number: 97-159 28869,29253,29607,'and 28980, Between July 1994 and February 1996, several design changes were made to the Plant Service 1 Water (WS) System requiring revision to the Updated Safety Analysis Report (USAR). The changes consisted ofinstallation of a split-type mechanical seal on the service water pump IWS0lPA and a %-inch drain line and valve (Engineering Change Notice (ECN) 28161). Two mechanical seal drain lines on this pump had a common connectiori that was disconnected and both lines terminated above the newly added drain funnel attached to the downstream portion of the same line (ECN 28788). The same changes were made to the other plant service water pumps 1WS0lPB and 1WS0lPC (ECNs 28789 and 28780). Finally, two high point vents and three
    ' drains were added to WS lines feeding component cooling water system heat exchangers
      , ICC01 AA and 1CC01 AB (ECNs 28869,-29253 and 29607). All changes affect piping shown in USAR Figure 9.2-1.
The WS system is not assumed to function following a design basis accident. USAR Section 9.2.1.1 indicates that WS is not required to ensure safe shutdown. Described changes are located such that their failure cannot impact any existing plant safety related components. If a functional failure of a component within the scope of these ECNs occurred, that failure will not prevent any
     . safety related system from performing its function.- A plant service water pump seal failure would only impact non-safety related screenhouse equipment; and is not considered a new failure. The new vents and drains are small in size (1/2- to 3/4-inches) and would not impact the 12- or 20-inch piping in which they are installed. Based on the minor scope of these changes, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR, nor does this activity create an
                                                 ~
accident or malfunction of equipment important to safety of a different type than previously evaluated. _ The WS System does not contribute to any acceptance limits within the CPS Technical Specifications.' Therefore, this change does not reduce a margin of safety as defined in the basis    .
for any Technical Specification.
A

t; 1.-                                   4                                                  Attachment 2 to U-603276 Page 40 of 252 3.8.1 and 3.8.2. Section 9.4.5.1 of the Updated Safety Analysis Report (USAR) discusses VD L              component failures. . The procedure change does not impact performance or operation of the l              system. Therefore, this change does not !acree the probability or consequences of a malfunction
             ! of equipment important to safety previously evaluated in the USAR. This change does not -
involve any physical modification of the plant, it dces not alter the method of operation of any
             - system, nor does it involve any postulated accident initiation event. Therefore, this change does      -
not increase the probability or consequences of an accident previously evaluated in the USAR.
This change does not affect the performance 'or configuration of the VD system.' Prior to exceeding design temperatures, VD system and diesel generators are secured. The DG is declared
: i.           ' inoperable before reaching the high temperature. Therefore, this change does not create an l:             accident or malfunction of equipment important to' safety of a different type than previously evaluated in the USAR. This change ensures that if DGs are not considered operable at high/ low
: outside temperatures, then appropriate TS LCO 3.8.1 and 3.8.2 actions are taken. Therefore, this procedure change does not reduce a margin of safety as defined in the basis for any Technical l              Specification.
l ENGINEERING EVALUATION OF INCREASE IN HORSEPOWER REQUIRED FOR HG l              COMPRESSORS Document Evaluated: Engineering Evaluation for'                                Log Number: 97-163 l                                        CR l-97-07-105 l
The engineering evaluation for Condition Report (CR) 1-97-07-105 determined the horsepower required to drive the Containment Combustible Gas Control (HG) mixing compressors under various accident operating conditions was different than had been used in the most recent ;
t               electrical voltage drop calculations and shown in Table 8.3-13 pf the Updated Safety Analysis           i Report (USAR) . Calculations show that the HG mixing compressors would require a maximum j
L               of 57 horsepower during any postulated accident scenario. The increased horsepower was added            l l               and the results indicated there is no impact on the Engineered Safety Feature equipment. The l               engineering evaluation used data different than that listed in USAR Table 8.3-13. Even using the        l
             - higher horsepower values, the maximum continuous rating of each Diesel Generator (DG) is not exceeded. The temperature in the HG compressor rooms increase negligibly due to the increase in horsepower.
Neither the function nor the 'way the equipment is being operated has changed, it is still within its design capability. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the USAR.
The capability'of the DG and HG mixing compressors is not exceeded and the increase in room L
temperature is within the equipment qualifications. Therefore, this activity does not create an accident or malfunction of equipment important to safety different than any previously evaluated lin the USAR. The purpose of USAR Table 8.3-13 and USAR Section 8.3.1.2.2 is to show that the maximum Loss-Of-Coolant Accident loeds do not exceed the capability of the continuous
: rating of the DG. The engineering evaluation concluded that the DG loading capability is not i
B u
L

Attachment 3 to U-603276 H                                                                                       Page 39 of 252 USAR Revision 7 reported that ECN 28980 had not been implemented as of the cutoff date for
        ' that reporto Since that time, ECN 28980 has been cancelled.
USAR DRAWING CHANGE ASSOCIATED WITH RCIC ROOM FLOOR DRAIN SUMP Document Evahiated:' USAR Change 8-003 '                                      Log Number: 97-160        .
Piping and Instrumentation Drawings (P& ids).M05-1070 and M05-1047-6 are being revised to show the inputs to the Reactor Core Isolation Cooling (RCIC) room. floor drain sump.
M05-1070 is associated with Updated Safety Analysis Report (USAR) Figures 6.7-1 and 3.6-1, sheet _54, and M05-1047-6 is not referenced in the USAR or appear as any of the USAR figures.
         -This drawing change removes the construction references and the nomenclature associated with              :
        . floor d' iains and the sump and replaces them with arrows directing the user to the appropriate Floor Drain System P&ID. USAR Section 6 7.2.3d describes the function of these drain lines and their respective check valves. This USAR change does not affect the function, configuration, or nwcrials of the systems in any way. Therefore, this activity does not increase the probability or cc        aences of a malfunction of equipment important to safety previously evaluated in the US AIL Since the affected systems will not be operated any differently due to the USAR change,
        ' the proposed will not increase the possibility or consequences of an accident previously evaluated or create a potential accident or malfunction of equipment important to safety of a different type.
This USAR change does not impact any Technical Specification safety limit, safety system setting, or limiting condition for operation nor the Bases for those requirements. Therefore, this activity does not reduce a margin of safety as defmed in the basis for any Technical Specification.
DIESEL GENERATOR HVAC SYSTEM OPERABILITY AT HIGH/ LOW TEMPERATURE Document Evaluated: OD & CR l-97-07-250                                       Log Number: 97-161 The Diesel Generator Heating, Ventilating, and Air Conditioning (HVAC) System (VD) design is based on a maximum outside temperature of 96 F and minimum outside temperature of +5 F.
The VD HVAC system is designed to limit the Diesel Generator (DG) room temperature to 130*F at a maximum outside temperature of 96.*F. If the maximum outside air temperature is exceeded, then the DG room temperature could go above 130*F. Condition Report (CR) 1-97-07-250 was initiated to identify concern with exceeding maximum and minimum outside temperatures. Previously, CR 1-97-06-302 established a minimum outside temperature design limit of +5 *F. Engineering calculations indicate that Division I DG room can maintain a temperature of 130 F at a maximum outside temperature of 102.7 F, and in Division 11 and III DG room at a maximum outside temperature of 104 *F. Therefore, a procedure change is being made to initiate compensatory action when outside temperatures are at or above 102.7*F and 104 *F.:
Clinton Power Station (CPS) Procedure 9000.01D001&2, " Control Room Surveillance Log,"
ensures proper control of DG room temperature by monitoring temperature and taking appropriate action per Technical Specification (TS) Limiting Condition for Operation (LCO)

    .                                                                                      Attachment 2 to U-603276 Page 42 of 252
        . Station (CPS) procedure 3214.01. The USAR states that service air will be supplied to plant components at 120 psig, when actually it is supplied and controlled at 110 psig. Table 9.3-1 indicates that Automatic Depressurization System (ADS) Air Amplifier inlet pressure is 80 psig, when actually it is regulated at 60 psig. Section 9.3.1.2.b an'd Table 9.3-2 states the ADS instrument air amplifiers boost the pressure to a minimum of 150 psig, however, minimum pressure is directly related to inlet pressure to the amplifiers. Normally, an operating pressure of  -
160 psig is maintained by the air amplifiers.
A low pressure alarm is provided to alert the operator to a malfunction of the normal instmment air supply to the ADS accumulators so the safety related backup supply can be remote-manually actuated. This change does not affect the safety related air supply to ADS or the ADS
        . accumulators. Therefore, this change does not affect the probability or consequences of a -
malfunction of equipment important to safety evaluated previously in the USAR. USAR Section             ,
15.2.10 analyzes the Loss ofInstmment Air System. Normal operating conditions are reflected in         i these changes and do not affect the probability or frequency ofinitiating events. Therefore, the probability or consequences of an accident previously evaluated in the USAR are not changed.
The changes are text changes and do not require a modification to plant equipment. The affected equipment can be expected to operate satisfactorily and air operated equipment necessary for safe shutdown of the plant will still fail in the safe direction. Therefore, the proposed changes do not create the possibility of an accident or a malfunction of equipment important to safety of a            1 different type than previously evaluated in the USAR. All affected equipment still meet the Technical Specification requirements for minimum air pressure in the ADS accumulators;
  .       therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
FREEZE SEAL ON FIRE PROTECTION SYSTEM LINE OFP66A-6 Document Evaluated: MWR D71467 Freeze Seal                                       Log Number: 97-173 In order to repair Standpipe Isolation Valve 0FP162F, per Maintenance Work Request D71467, a freeze seal is required on line 0FP66A-6. This is a six inch line that feeds 14 hose stations on all levels of the Control Building. With the exception of the 14 hose stations, the fire Protection (FP) system will remain operable during the freeze sealing activities. The freeze seal will be performed per Clinton Power Station (CPS) procedure 8208.01," Freeze Sealing." The hose                   i stations will be restored following the completion of the repair activities on valve 0FP162F. This freeze seal is considered a change to the facility as described in the Updated Safety Analysis Report (USAR).'
Analysis determined the worst case flood would occur from a break on 24 inch line 2WS02F.
This analyzed line break would cause no damage to safe shutdown equipment. 'If the freeze seal on 6 inch line 0FP66A-6 were to fail with the valve disassembled, the leakage rate would be less L         than the analyzed leakage rate from 24 inch line 2WS02F. Therefore, the freeze sealing activities
;         on 6 inch line OFP66A-6 will not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. Also, compensatory measures l                          .

e Attachment 2 to U-603276                 l Page 41 of 252              l exceeded; therefore, this activity does not reduce a margin of safety as defined in the basis for any     i Technical Specification.
CONSERVATIVE DECISION MAKING ALLOWANCE FOR DIESEL GENERATOR HVAC SYSTEM OPERABILITY Document Evaluated: PDR to CPS Procedure                                      Log Number: 97-168 9000.01D001/D002 Updated Safety Analysis Repon (USAR) identifies extreme temperatures for the site based on surrounding areas of 112 F and -22 F. Condition Repon (CR) 1-97-07-250 identified that the Diesel Generator (DG) Heating, Ventilation, and Air-Conditioning (HVAC) system (VD) is
                                                                                                            ]
designed for a temperature range of-2 F to 96 F based on the 1% and 99% extremes as recommended by ASHRAE. Based on evaluations for CR l-97-07-250 and CR l-97-06-302, the VD system operability may be in question at temperatures below 5 F and above 103 F.
Calculations indicate that with outside temperatures below 104 F, the VD system will maintain temperature in the DG rooms to less than 130 F. This change provides a way to determine DG and VD operability in accordance with Clinton Power Station (CPS) procedure 1014.06,
  " Operability Determination (OD),"instead ofimmediate entry into a Technical Specification I
shutdown action statement.
Requiring an OD per CPS 1014.06, so that appropriate actions and compensatory measures are taken, enhances safety. Performing an OD does not prevent shutdown of the plant, but allows appropriate informed conservative decisions based on the best available information. This change does not involve any physical modification of the plant, does not alter the method of operation of any system, nor allow operation inconsistent with Technical Specification requirements.
Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Repon (USAR). The operation >f the VD system for all months of the year except July (without operator compensatory action) will be within the Design / Licensing Basis. Funher evaluation is ongoing and is expected to show full qualification to the room temperatures even at the July extremes. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. This change does _not create the possibility of an accident or malfunction of equipment imponant to safety of a difTerent type than previasly evaluated in the USAR because it does not change the basic VD or DG system performance or configuration.                   i Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical     l Specification.                                                                                            j SERVICE AIR PRESSURE AS STATED IN THE USAR Document Evaluated: USAR Change 7-303                                         Log Number: 97-170 Updated Safety Analysis Repon (USAR) Sections 9.3.1.2.a and 9.3.1.2.b and Table 9.3-1 are inconsistent with the System Description for Service Air /Instmment Air and Clinton Power

e
        .-                                                                                      Attachment 2
'                                                                                               to U-603276 Page 44 of 252 The potential for.each seal has been evaluated ~as having no potential for draining the reactor
              . vessel. Each floor drain pump has a capability of pumping up to 120 gallons per minute. If the freeze seals fail, the most leakage that will occur would be approximately 20-22 gallons per
              .' minute. Therefore, the probability or consequences of a malfunction of equipment important to safety evaluated previously in the Updated Safety Analysis Report (USAR) is not increased. The installation or removal of these freeze seals will not affect any radiological parameters identified in the USAR while in Mode 4; therefore, the probability or consequences of an accident previously
              . evaluated in the USAR is not increased. If the freeze seals fail, the leakage would be less than a
[
l               feedwater line rupture and within the normal makeup water means. Therefore, this accident could         ,
l               not create the possibility of an accident of a different type than previously evaluated in the USAR.
              - If the freeze seals fail, the maximum leakage would be approximately one inch ofwater covering 1 the Drywell basement floor. The Floor Drain pumps would remove the water and the Liquid
              ; Radwaste systems would process the water. Since no safety related equipment would be affected by the water, the installation of the freeze seals would not create the possibility of a malfunction l     ,
of equipment important to safety of a different type than previously evaluated ir, the USAR. The installation or removal of the freeze seals does not constitute an operation to potentially drain the L               reactor vessel and Reactor Water Cleanup will act as an alternate shutdown cocling method.
Therefore. this activity does not reduce a margin of safety as defined in the basis for any Technical
!.              Specification.
L
!               TEMPORARY POWER TO STATION ALARM PANEL                                                                   I Document Evaluated: Temporary Modification 97-0 3                               Log Number: 97-177 r
Temporary Modification 97-073 provides temporary 120 VAC power to the Merge-Isolate / Alarm L               panel located on 781' elevation Control Building and allows the station alarms near the Remote L Shutdown pan'   e l to function properly. The panel lost power when the Division 1 electrical bus was de-energized. The temporary modification will lift and tape the wires originating from the normal power source and install temporary power fro a a local 120 VAC outlet. The temporary
              - power cord u' sed will consist of a standard extension cord either modified to allow termination to a terminal board, or connected to a pigtail connected to the terminal board. The amperage load on the cord is 'within the specified range of the utility outlet and does not impact any load calculations.J
               ~ The Merge-Isolate / Alarm panelis not safety-related and performs no safety function. The utility outlet is within 5 feet of the panel and the cord will be properly secured in the routed path. The l                addition of the temporary power cord will not increase the probability of a malfunction of
[
equipment important to srfety previously evaluated in the Updated Safety Analysis Repon (USAR). Due to the short run and the fact that the cord will be secured, the temporary power cord will'not interfere with the operation of the Division 3 equipment located nearby. Therefore, this change does not increase the consequences of an equipment malfunction important to safety
               ? previously evaluated in the USAR. Providing temporary power to the panel cannot increase the i
probability or consequences of any accidents previously evaluated, because no previous accident scenarios considered this panel. The panel normally has a power source, and the normal g

Attachment 2
                            .                                                        to U-603276
* Page 43 of 252 for each hose station taken out of service will be taken per CPS procedure 1893.01," Fire Protection Impairment Reporting." Therefore, the changes will not increase th'e possibility or
                                                                                                                ]
consequences of an accident previously evaluated in the USAR. The accidents and equipment malfunctions possible due to freeze sealing activities of the FP line are caused due to internal flooding or lack of fire protection. Internal flooding is evaluated in the USAR and the compensatory actions ensure the protection provided by the FP system is not corrpromised.                 .
Therefore, the freeze sealing activities will not create the possibility of additional equipment malfunctions of a different type than any evaluated previously in the USAR. The FP system is not an assumed factor in establishing a margin of safety for any Technical Specification; therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
FEEDWATER HEATER VENT AND DRAIN VALVES IN DIFFERENT CONFIGURATION THAN SHOWN IN THE USAR Docu ent Evaluated: USAR Discrepancy Related to                                 Log Number: 97-175 CR l-97-02-032 Piping and hstrumentation Drawing (P&lD) M05-1004 inaccurately reflects the as-built configuration of a section of non-safety related piping. The P&lD for drain and vent valves at the outlet of feedwater heaters 6A and 6B shows a symbol of a double-block globe valve, when actually, it is two globe valves in series. The Updated Safety Analysis Report (USAR) Figures associated with this P&lD are Figures 10.4-7 and 3.6-1, sheet 14. Functionally, the configuration existing in the plant is the same as the double block valves shown in the USAR. The USAR does not give the drain and vent line valves credit for mitigating radioactive releases from Clinton Power Station. Therefore, the probability or consequences of an accident or a malfunction of equipment important to safety evaluated previously in the USAR does not increase. The only system the drain and vent Sne valves affect is the Feedwater System and the USAR already considers Loss of Feedwater. This change will not create an accident or malfunction of
   . equipment important to safety of a different type than previously evaluated in the USAR. The plant configuration is functionally the same; therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
FREEZE SEALS FOR RWCU RECIRCULATION LOOP B SUCTION VALVE IG33F106
   ' Document Evaluated: MWR D34974 for Freeze Seals                             Log Number: 97-176 R/l Maintenance Work Request (MWR) D34974 installs two freeze seals to repair leaking packing on valve IG33F106, Reactor. Water Cleanup (RWCU) recirculation loop B suction valve. One freeze seal will be installed on the upstream side of the valve on stainless steal line 1RT01EA4 and the second freeze seal will be installed on the downstream side of the valve on carbon steel line IRT01 AA4. The freeze seals will be installed to assist maintenance activities and allow safe replacement ofvalve packing for IG33F106. The freeze seals will not adversely affect the piping and will be installed per Clinton Power Station procedure 8208.01, " Freeze Sealing."

a
   ~
F                                                    ' Attachment 2 d     .
to U-603276 -
t                                                                                        Page 45 of 252 operation of the panel does not change. Therefore, the temporary modification does not create an accident of a different type. The panel will not fail in a different way than was possible before and the temporary cord will not influence the operation of the Division 3 equipment. Therefore, the
          ? temporary modification cannot create a malfunction of equipment important to safety of a
: different type. The temporary modification does not cause any structure, system, or component
          . to operate outside ofits design basis; therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Spec;ification.
          - CONTROL BUILDING EL' EVATION 751'-0" HIGH TEMPERATURE DETECTION Document Evaluated: Modification VL-008                                          Log Number: 97-179 Laboratory Heatins, Ventilation, and Air-Conditioning (VL) System steam lines are classified as high energy lines since the operating temperature in these lines is in excess of 200 F The Control Building 737' elevation corridor and 751' e'evation mezzanine floor could exceed its design temperatures if this piping failed.' The VL Humidification system is non-safety related, non-seismic, and non-Class IE. The purpose of Modification VL-008 is to add temperature switches to shut off the VL humidification unit upon detection of high area temperature. The modification adds safety related temperature switches at ' elevation 751' where the humidification unit and steam lines are located.f Non-safety related power supply will support the temperature switches, relays, push-button switch and terminal blocks. Upon loss of the non-safety re:ated-          j power supply, the temperature switches will shut off the VL boiler. High area temperature and
          - isolation of the boiler will be annunciated at local panel.
Rupture analysis of this piping was not performed as Updated Safety Analysis Report (USAR)
Section 3.6 assumes a loss of offsite power coincident with the line break. However, availability of offsite power should be assumed based on USAR Section 3.6.1..l.4.g. Calculation EMD-053176 indicates blowdown force from postulated VL steam lines does not have sufficient energy to require consideration forjet impingement or pipe whip. . To ensure temperatures stay L within' design limits, safety related temperature switches are installed at 737' and 751' elevation
          . Cont'r ol Building per USAR Section 8.3.1.1.d. The setpoint of the temperature switches are set I
below maximum allowable temperature. Therefore, this change will not incmase probability or consequences of a inalfunction of equipment important to safety evaluated previously in the USARL The area environment of the Control Building elevation 737' and 751' is required to stay the same'during normal plant operating conditions and accident conditions. Addition 6f the temperature switches supports this design by shutting off the VL boiler in the event of a steam line break. Therefore, this modification does notjmpact the probability or consequences of an accident previously evaluated in the USAR. Relays and terminal blocks installed in the new panel are seismically mounted, the function or performance of the equipment is not impacted. The VL system itselfis non-safety related and not required for the safe shutdown of the plant.
                                                                 ~
Considering this, the modification will not create a new type of accident or equipment                  i malfunction. Temperatures in the area stay within design limits, which ensures that safety related
           . equipment is uncha:.;;ed. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification l

Attachment 2 to U-603276 Page 46 of 252 INSTALL ISOLATION VALVE IN RETURN LINE TO CONDENSER FROM PROCESS SAMPLE PANEL 1PL22J
  ' Document Evaluated: ECN 29543, Change to USAR .                               Log Number: 97-181 Figure 9.3-4 Sheet 5 -
Engineering Change Notice (ECN) 29543 installs isolation valve IPS121 in line 1PS42A-2 to allow isolation of the Process Sampling (PS) system Condenser Hotwell Tube Tray Sample Panel (1PL22J) from the Main Condenser.' In conformance with Updated Safety Analysis Report
  . (USAR) Section 9.3.2.4, 'the isolation valve is a normally open valve and provides for isolation only when maintenance is necessary. The addition of the isolation valve allows maintenance on .
the sample panel eductors when the unit is on-line.
Sample panel IPL22J is located in the basement of the turbine building and does not interface with safety-related equipmenty For this reason, the change does not affect the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The only accident scenario possible due to maintenance performed on the panel with the
,   unit on line is a loss of condenser vacuum event. Two possibilities exist: the packing may degrade in which case the leakage would not be great enough to cause a loss of condenser vacuu.m trip, and the other would be a tag out error in which case the error would be detected on initial breach and corrected prior to a trip occurring. Therefore, this change does not significantly increase the probability of an accident previously evaluated in the USAR. Should a postulated loss of condenser vacuum occur, the method of progression is unaffected and, hence, any radiological release would be as' evaluated. Therefore, this activity does not increase the consequences of a previously evaluated accident. The change can only credibly impact a loss of condenser vacuum trip which is previously evaluated. Therefore, the change does not create the                '
possibility of a new type of accident. The possibility of packing leakage, disc leakage, and operational error exists, all of which are evaluated in the USAR; no new type of malfunction is introduced.. The portion of the PS system affected by this change is not governed by the Technical Specifications; and therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REMOVAL OF DOORS ON DIVISION 3 DG CONTROL PANEL lE22S001B Document Evaluated: Temporary Modification 97-076 '                          Log Number: 97-184
   ' During an investigation for Condition Report 1-97-07-250, it was determined that temperatures in the Division 3 Diesel Generator (DG) Control Panel (IE22S001B) could be 23 F higher than the 130 F design temperature in the room. Two components inside the panel are only qualified for a maximum ~ ambient temperature of 140 F. The failure of these two components could result in the failure of other safety-related components. Temporary Modification 97-076 removed the panel doors on IE22S001B to allow air circulation through the panel, which will ensure the temperature inside the panelis maintained below 140 F. The removal of these two doors impacts the panel description in the Updated Safety Analysis Report (USAR).
1 1

    .;                                                                                             Attachment 2 to U-603276 Page 47 of 252
               ' The panel doors provide protection against dust and water. The room ventilation system has
               > filters to remove dust and particulates in the air stream and the Facilities Group maintains the cleanliness of the roomM Another concern is the wetting of compor.ents important to safety if a
                ' leak was to occur. ! The components inside the panel are not in the line-of-sight of the water lines
               - and most of the components are enclosed so that indirect water splashing would not impact operation. No forceful spray or impingement sources are in close proximity to the cabinet. The         .
                ; seismic qualification has been evaluated as acceptable with the doors removed. Therefore, this J change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. USAR Chapter 15 does not address the failure of a DG as an initiating' event. The loss of one diesel generator system and its associated load group will not prevent safe shutdown of the reactor following a loss-of-coolant accident and loss of offsite power. Therefore, this change does not impact the probability or consequences of an accident previously evaluated in the USAR. ' Because the failure of one DG is not an initiating
                < event for any accidents and the High Pressure Core Spray power supply are not affected by this
                ' change; the removal of the' panel doors do not create an a'c cident or a malfunction'of equipment important to safety of a different type. . The DG will start and operate at its rated values. The change does not affect the critical DG parameters described in the Technical Specifications.
Therefore; the removal of the panel door does not reduce a margin of safety as defined in the basis
                 . for any Technical Specification.
NEUTRON SOURCE REMOVAL FROM THE REACTOR VESSEL -
_. Document Evaluated: USAR Change 8-001                                           Log Number: 97-185
                 . Updated Safety Analysis Report (USAR) Change 8-001 revised several USAR passages to
                . correct discrepancies identified in Co4 tion Report (CR) 1-97-05-343 and to reflect the rc aval of the neutron sources from the~reacto, se. These' sources had been removed durin'        g the arst refueling outage'(RF-1).and were previously reported under Safety Evaluation 89-0040.
                                  ~
The neutron sources functioned as a source of neutrons for initial unit startup. These sources facilitated neutron flux monitoring and observation of flux level changes within the reactor.
: Intrinsic neutron sources now provide that function. These sources and their associated function are not associated with the podulated initiators of any evaluated accident. . The source function provided by these devices does not affect the sequence of events or propagation of any evaluated event; Nor do.the' sources provide any mitigatory function during any of accidents evaluated in
    ,               the USAR. Therefore, the removal of the neutron sources does not increase the probability or consequences of a previously evaluated accidentf The sources are physically separated from equipment important to safety with the exception of the fuel cladding. The absence of the sources does not introduce a degradation or failure mechanism not already evaluated in the core design.
                 ' The neutron sources do not provide a mitigatory ftmetion upon failure of equipment important to
                  . safety; nor do the exacerbate the postulated consequences of a failure of existing equipment.
Nuclear instrumentation sensitivity is sufficient to compensate for the removal of these sources.
Thus, no postulated impact on the ability of the operator to determine and interpret core neutron      ;
o-                  flux levels is present.- Therefore, the removal of the neutron sources does not increase the i
J

1
                                      .e                                                                        ,
                              -U                                                          Attachment 2
                                      .,i-                                                                                       ,
to U-603276                    -
Page 48 of 252 probability or consequences of a malfunction of equipment imponant to safety previously evaluated in the USAR. . The absence of the sources does not introduce any analytical or physical                         i interaction mechanism that could initiate a new type of accident or malfunction of equipment important to safety not previously evaluated in the USAR. Safety analyses assumptions regarding neutron flux and reactor power level are unaffected by this activity. Therefore, the removal of the sources does not include the potential for a reductio'n in a margin of safety as dermed in the bases of any Technical Specification.
CHANGE RELIEF VALVE ICC318 SETPOINT :
                                 ~
    ' Document Evaluated: ECN 30271                                                      - Log Number: 97-186 The current 'setpoint for relief valve 1CC318 is 200 psig; Engineering Change Notice (ECN) 130271 replaces this' valve with a similar relief valve with a setpoint of 500 psig. Actual operating -
pressure in Component Cooling Water system (CC) can be up to 120 psig with pressure over 200 psig possible during system transients. These transient values are close to the current relief valve setpoint;- The relief valve.will be set at 500 psig and will discharge into the containment atmosphere if the CC header pressure exceeds 500 psig during a postulated Loss-of-Coolant
     ' Accident (LOCA). The setpoint change assures that the non-safety related piping does not adversely affect the associated safety related piping attached to it. This change affects Updated Safety Analysis Report (USAR) Figure 3.6-1, Sheet 36 and Figure 9.2-3, Sheet 3.
The increase in the setpoint is stillw  ' ithin the design pressure limits of the piping. The change serves to decrease the likelihood that the valve will lift during minor operational challenges.
Therefore, the relief valve setpoint change does not increase the probability or consequences 'of a malfunction of equipment important to safety. During a LOCA-induced heatup and -
overpressurization of the piping, the relief valves assure that the non-safety related piping does not adversely affect the associated safety related piping attached to it; Therefore, the change does
    - not increase the probability or consequences of an accident previously evaluated in the USAR.
The piping has been evaluated for dynamic loads in containment resulting from seismic
    ; occurrences, Main Steam System Safety Relief Valve discharges, and LOCA and will not collapse and damage safety-related equipment. Therefore, the change does not create an accident or
    - malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The piping meets operating pressure ranges and will not adversely affect associated safety related piping; thus, this activity ldoes not reduce a margin of safety as defined in the basis for any
     --Technical Specification.
CHANGE RELIEF VALVE ICC319 SETPOINT
    ~. Document Evaluated: ECN 30272                                                      Log Number; 97-187 The current setpoint for relief valve ICC319 is 200 psig; Engineering Change Notice (ECN)
      .30272 replaces this with a sinalar relief valve with a setpoint of 500 psig. Actual operating pressure in Component Cooling Water System (CC) can be up to 120 psig with pressure over 200 m
t

    '*                                                                                       Attachment 2 to U-603276 Page 49 of 252 psig possible during system transients. These transient values are close to the current relief valve setpoint. The new relief valve'will be set at 500 psig and will discharge into the containment
           . atmosphere if the CC header pressure exceeds 500 psig during a postulated Loss-of-Coolant Accident (LOCA). The setpoint change assures that the non-safety related piping does not adversely affect the associated safety related piping attached to it. This change affects Updated Safety Analysis Report (USAR) Figure 3.6-1, Sheet 36 and Figure 9.2-3, Sheet 3.                        -   !
The increase in the setpoint is still within the design pressure limits of the piping. The change serves to decrease the likelihood that the valve will lift during minor operational challenges.
Therefore, the relief valve setpoint change does not increase the probability or consequences of a
            ~ malfunction of equipment important to safety. During a LOCA-induced heatup and overpressurization of the piping, the reliefvalves assure that the non-safety related piping; does
       . not adversely affect the associated safety related piping attached to it. Therefore, the change does not increase the probability'or consequences of an accident previously evaluated in the USAR.
The piping has been evaluated for dynamic loads in containment resulting from seismic
;            occurrences, Main Steam System Safety Relief Valve discharges, and LOCA and will not collapse              C and. damage safety-related equipment. Therefore, the change does not create an accident or                 l malfunction of equipment imponant to safety of a different type than previously evaluated in the USAR. The piping meets operating pressure ranges and will not adversely affect associated safety related piping; thus, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
AREA, EQUIPMENT, AND PERSONNEL CONTAMINATION CONTROL Document Evaluated: Change to USAR Section 12.5.3.4                            Log Number 97-191
           - This change adds provisions for temporary Radiological Control Area (RCA) egress points with the approval of the designated Radiation Protection Manager. Section 12.5.3.4 of the Updated
           ' Safety Analysis Repon (USAR) describes the location of the two established RCA egress points, the addition of a condition to add temporary egress points necessitates a USAR change. This change does not affect any systems, structures or components. In addition, this change does not
           ; impact any content of procedures which operate, maintain, or test equipment. Therefore, this
           . change does not increase or create the probability or consequences of a malfunction of equipment important to ' safety previously evaluated in the USAR. This change does not impact requirements of Title 10 Part 20 of the Code of Federal Regulations for exposure limits, exposure monit'oring, or As-Low-As-Reasonably-Achievable Program requirements. Therefore, this change does not increase or create the probability'or consequences of an accident previously evaluated in the
           . USAR. This change does not impact any Nuclear Regulatory Commission acceptance criteria or any plant equipment. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
4

Attachment 2 to U-603276 Page 50 of 252 f
ELECTRONIC TEMPERATURE CORRECTION OF BATTERY ELECTROLYTE SPECIFIC GRAVITY MEASUREMENT
     - Document Evaluated: Bases Change Request BE-97-016                                Log Number: 97-198 The Technical Specification Bases change will allow for an appropriate temperature correction             -
method for battery electrolyte specific gravity measurements. The present method of temperature correction is a manual correction applied to reading a float-:ype hydrometer. The change allows use of a new digital hydrometer with greater accuracy and an automatic, electronic temperature correction. As required by the current . Technical Specification Bases for Surveillance Requirements 3.8.6.1 and 3.8.6.2 and Table 3.8.6-1, the digital hydrometer will be referenced to 77 F. -
The change to the Technical Specification Bases only allows an alternate method of correcting specific gravity measurements; it does not change the criteria by which the battery is maintained.
    ' The change improves the accuracy of battery electrolyte specific gravity measurements and consequently indicates a more accurate condition of safety related equipment. Therefore, the
     . probability or consequences of a malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). The change does not change the manner in which the batteries are operated, installed, or maintained and makes no change to the design calculations. Therefore, this change will not increase the probability or consequences of an accident previously evaluated in the USAR. This change does not create a new accident scenario or malfunction of equipment of a different type than previously evaluated in the USAR, because neither plant equipment, nor the operation of plant equipment changes. The method of
    - temperature correction changes, not the Battery Cell Parameter Requirements as prescribed in Table 3.8.6-1 of the Technical Specifications. As a result of the method change, the digital hydrometer will be more accurate than the manual method. Consequently, this activity does not
     ' reduce a margin of safety as defined in the basis for any Technical Specification.
    ' FIRE PUMP FUEL SYSTEM MODIFICATIONS Document Evaluated: Modifications FP-102 and FP-103
                              .                                                         Log Number: 97-199 Modifications FP-102 ad FP-103 make numerous changes to the fire pump diesel engine fuel supply system. These changes increase the fire protection pumps' availability and reliability.
These changes include the following: replacing copper materials in contact with Fire Pump 0FP0lPA/PB diesel engine fuel with stainless steel; adding 24 inch manways to the top of the fuel i storage tanks to allow coating of the tank interiors; sloping the tanks toward the drains to protect the tank discharge to the diesel fuel pumps and allow for more accurate sludge sampling through                .
the drain; replacing the simplex diesel fuel filters with duplex diesel fuel filters to allow filter          J change out while the engines are in operation; and adding fuel filters to the fill lines to minimize        -
tank contamination by fill line sludge and particulates.
i a
e

I   -
    -                                                                                               Attachment 2 to U-603276 Page 51 of 252 Changing the material from copper to stainless steel reduces "geiling" effects on the fuel, thereby reducing the potential for clogging the fuel line and fire pump inoperability. Sloping and coating the tanks rv ce tank corrosion and fuel contamination and reduces potential for fuel line clogging and Fire Pump inoperability. The coatings have been demonstrated to possess excellent adhesion propertier rnd thus, are not probable sourcet of contamination for the fuel oil. The new duplex fuel oil filter is less likely to cause diesel inoperability because the fuel flow can be               .
rerouted through the companion filter if one filter becomes clogged. The addition of the fill line filters reduces the contamination reaching the Fuel Oil Storage Tanks, thereby reducing the potential for fuel line clogging and fire pump inor rability. The modifications increase the fuel system reliability by climinating some fuel fouling sources, and providing redundant fuel oil filtration paths. Theref;ca. these changes do not increase the probability or consequences of a malfunction of equirrnent i.nportant to safety previously evaluated in the Updated Safety Analysis Report (USAR). The fire pump Diesel Engine fuel system is not involved in the initiation of any accidem. Therefore, these changes do not increase the probability or consequences of an accident prev.ously evaluated in the USAR. These changes do not create any new hazards, flow paths, or interfaces.. Therefore, these changes do not create the possibility of an accident of a different type than previously evaluated. The fuel storage tanks' fill line filters are equipped with an automatic bypass feature which allows' the tanks to be filled in an emergency even if the fill line filters are clogged. The duplex filter is designed to provide complete flow diversion from one side filter to the other without interrupting fuel supply to the diesel engines. Since interfaces with equipment important to safety have not been changed by this modification, the change does not create the possibility of a new type of equipment malfunction. These modifications do not adverselp affect the capability of plant systems to achieve and maintain safe shutdown in the event of fire.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
FIEEZE SEAL INSTALLATION FOR CYCLED CONDENSATE PIPING HEADER Document Evaluated: MWR D64257                                                              Log Number: 97-200 In order to isolate the Containment Cycled Condensate (CY) header from the Auxiliary Building CY header, Maintenance Work Request (MWR) D64257 installs a freeze seal upstream of valves ICYO15 and ICY 016. The freeze sealis necessary to rework valve ICY 015 that is leaking past its seat. The equipment serviced by the CY in the Containment and Drywell will be out of service while the freeze seal is installed, with the possible exception of the Reactor Water Cleanup System filters. The Auxiliary Building CY header remains in service while valve ICY 015 is being reworked due to the location of the freeze seal.
Three safety-related Containment Isolation Valves (CIVs)in the Auxihary Building are in the immediate vicinity of valve ICY 015 and may be affected if the freeze seal fails. Per Updated Safety Analysis Report (USAR) Section 3.11.9.19, all three of the CIVs are EQ qualified to 100%
humidity and each has been qualified fc. condensation and steam due to a moderate energy line break. Therefore, low energy water spray from a freeze seal failure will not increase the probability or the consequences of a malfunction of equipment important to safety evaluated

Attachment 2 to U-603276 Page 52 of 252 previously in the USAR. Per USAR Section 9.2.6.1.1, the CY system provides water to safety related systems. However, the CY system does not perform any safety related function itself. No credible accidents exist that would be affected by the failure of the freeze seal on this CY line.
The Flood Analysis Report covers the impact of flooding if the freeze seal fails while the 6" valve was open to the atmosphere. The Flood Analysis Report came to the conclusion that the maximum flood level will not affect any equipment required for safe shutdown. ThcVore, the possibility of an accident or malfunction of equipment important to safety of a differeat type than previously evaluated does not increase. The change does not affect any protective barriers or                                                                      ,
equipment important to safety; tLefore, this activity does not reduce a margin of safety as                                                                        l defined in the basis for any Technical Specification.
REVISION OF REFERENCED DOCUMENTS CONTAINING CAFCOTE H AS FIRE                                                                                                      )
PROOFING MATERIAL l Document Evaluated: ECN 30405                                                                                                             Log Number: 97-202       ;
Engineering Change Notice (ECN) 30405 approves Cafcote 800 as an equal to Cafcote H as one l of the materials used to fire proof certain structural steel components. This change revises j Updated Safety Analysis Report (USAR) Appendix E Section 3.1.2.2.6 to state "or its approved i equal" when referring to Cafcote H. Since Cafcote H is no longer commercially available and its approved equal is not manufactured by the same company, the USAR text needs to be revised.
An approved equivalent product provides the same fire protection rating as Carcote H.
Therefore, this change does not increase the probability or consequences of a malftmetion of equipment important to safety previously evaluated in the US AR. This change does not affect initiating events, the ability of the Fire Protections System to function as designed, or the ability to acWve post-fire safe shutdown. Therefore, this change dc.es not increase the probability or consequences of an accident previously evaluated in the USAR. The fire protection rating of steel beams and columns requiring fire-resistive coating does not create the possibility of an accident or equipment malfunction of a different type. The Technical Specifications do not address the fire
, protection rating of structural steel beams and columns requiring fire-resistive coating nor the l specific fire-resistive coating material used as described in this change. Therefore, this activity j does not reduce a margin of safety as defined in the basis for any Technical Specification.
i
! REMOVE THE DRYWELL FLOOR DRAIN WEIR BOX FLOW MONITORING SYSTEM FROM THE TECHNICAL SPECIFICATION BASES Document Evaluated: CR l-97-09-102                                                                                                        Log Number: 97-204 The Leak Detection System provides sensors and instrumentation to indicate and alarm abnormal leakage from the Reactor Coolant Pressure Boundary (RCPB) and, in certain cases, initiate signals for automatic closure ofisolation valves external to the Drywell. Monitoring the Drywell floor drain flow rate, Drywell cooler condensate flow rate, and airborne gaseous and particulate radioactivity are the primary leak detection methods for small unidentified leaks within the Drywell. They are continuously recorded in the Main Control Room with appropriate alarms.
5

Attachment 2         j l  c
!                                                                                        to U-603276          ;
Page 53 of 252 l      The Drywell Floor Drain Flow rate is the only one that quantifies the leak rate. Currently, three alternate systems are available that are fully capable ofindependently meeting all functional and      l l
licensing requirements. This change removes the Drywell Floor Drain Weir Box Flow Monitoring System from the Technical Specifications (TS) Bases section of Limiting Condition for Operation 3.4.7 and updates the appropriate Updated Safety Analysis Report (US AR) sections. The other two flow monitonng systems are capable of performing their required functions independent of this system.
USAR Section 6.2 describes an accident involving a guillotine break of the recirculation system        )
suction piping involving leakage into the drywell. USAR Section 7.7.1.24.1 explains that no credit is taken for operation or operator reliance upon the leakage monitoring instrumentation.
The leak detection system continues to function as before, but now another more responsive             l altentate system increases system availability. The proposed change involves equipment that provides indication, it does not affect any initiators of any previously evaluated accidents.          I Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. Nor does this change increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The proposed change does not involve a change in the operation of the plant and it does not introduce any new failure modes. Therefore, this changc Joes not create an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The Leak                 i Detection System provides an early detection metliod for RCPB leakage. Leakage in excess of            l the TS allowable limits could result in continued degradation of the RCPB. The TS acceptance          )
limits for early identification of RCPB leakage remain unchanged. The TS Bases will be revised to eliminate any credit potentially tsken for the weir box instrumentation. Because the TS allowable limits are unchanged and because of the equivalence of alternate leakage detection          i systems, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
l USAR FIGURE 9.2-16 CHANGE                                                                              l Document Evaluated: CR l-97-08-085 and                                            Log Number: 97-205 USAR Change Package 8-014 Per the control valve specification sheet, this change revises Updated Safety Analysis Report (US AR) Figure 9.2-16 to show chilled water regulating valves OVS071 A and OVS071B fail in the closed position upon loss ofinstrument air. The original design and function of these valves remain unchanged. The closing of these valves isolates chilled water to the Service Building water chillers which are non-safety related.
Per original plant design, the " fail closed" position of the valves stops chilled water flow to the Service Building water chillers. USAR Sections 9.2.8.4.1.1,9.2.8.4.1.2(a), and 9.2.8.4.3 state the Service Building Chilled Water System is a non-safety related system. Therefore, the proposed change does not increase the probability or consequences of a malfunction of equipment imponant to safety previously evaluated in the USAR. The original design function of the valves

L L
* Attachment 2
                                                                                                                        +
to U 603276 Page 54 of 252
: remains the suae and the change has no effect on the continued operation of any safety related equipment' The failure mode of these valves and their associated system is not an initiator to any evaluated accident nor do they function to mitigate any evaluated accident.L Therefore, this Lchange does not increase the probability or consequences' of an accident previously evaluated in
          . the USAR.3The function of any system or component used to support safe operation of the plant remains unaffected? The change does not create an accident or a malfunction of equipment                      .
           'important to safety of a different type than previously evaluated in the USAR. Chilled water regulating valves' do not support and have no Technical Specification or Operation Requirements Manual requirements; therefore, this activity does not reduce a margin of safety as defined in the
          - basis for any Technical Specification.
FLOOR DRAIN AT 702; ELEVATION IN CONTROL BUILDING '                                        .
Document Evaluatedi CPS Procedure 1870.02,. .                                    .. Log Number: 97-209
                                        " Control of Drain Socks and Plugs" '
This activity allows for the installation'of a floor drain flow blocking device in the drain located at 7 702' elevation Control Building, approximately 10 feet from column 135-AA (DWG A30-1000-06A, C-5) per Clinton Power Station (CPS) procedure 1870.02, " Control of Drain Socks and
  .          Plugs." The installation of this device will stop oil and/or solvent intrusion into the floor drain .
system in accordance with CPS procedure 1870.02 requirements. The device is required as a
: result of activities per CPS procedure 8216.03, " Safety / Relief Valve Testing Inspection."
Eventually, without the floor drain plug being installed, incompatible fluids could migrate into the
           ' Radwaste Treatment System causing degradation in the evaporator, filter, demineralizer processing. The floor drain system's function of collecting liquid drainage from the local area of the 702' elevation Control Building is impacted by this change. This is contrary to the Updated Safety Analysis Report (USAR) Sections 1.2.2.8.10,9.3.3, and 12.1.2.1.5, which basically L describe the floor drain system as a system to collect radioactive and potentially radioactive waste liquids from their points of origin in the Drywell, Containment, Fuel Bldg., Aux. Bldg., Turbine Bldg., Radwaste Bldg , and Control Bldg., and to transfer them to the collecting vessels of the
           ;Radwaste Treatment System for processafor reuse or disposal. This is the only change identified to the floor drain function as described by the USAR.
If this local area drain is plugged per CPS procedure 1870.02, no credible scenario exists which
      .      would increase the probability'or consequences of a malfunction of equipment important to safety.
The failure of the floor drain system will not compromise any safety-related system or component and/or will not prevent the safe shutdown of the plant; therefore, no significant occurrence can be
           - postulated that would create the possibility of an accident or malfunction of a different tyne than previously evaluated. No credible scenario exists that would impact a margin of safety by the installation of the floor drain blocking device, since the floor _ drain system is defined as non-safety
          . related, supports no safety-related function, and is not relied upon through the Technical Specificat6ns or Operational Requirements Manual. ' Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
L

V a a 1
i V:                                                                                       Attachment 2 to U-603276 Page 55 of 252 RADWASTE DRAIN CONNECTIONS TO ENTRAINMENT SEPARATOR DRAIN Oocument Evaluated:- ECN 30263                                           Log Number: 97-210 Engineering Change Notice (ECN) 30263 corrects Updated Safety Analysis Report (USAR)
Figure 11.2-2, sheets 19 and 20. In these figures, two drain lines are incorrectly shown connecting down stream of the loop seal bypass from the chemical radwaste entrainment f separator. The two drain lines are a radwaste evaporator header drain line and a condenser vent gas cooler drain line. These two lines are still tied into the entrainment separator drain line         4
               ' downstream ofits loop seal.- The difference is that the connections are upstream of the loop seal       I bypass line instead of downstream. During plant operation, the bypass line is r.armally isolated to -
establish the loop seal. As a result, the revision of USAR Figure 11.2-2, sheets 19 and 20, to correctly show these two drain lines has no affect on the operation of the chemical radwaste l                         .
system.
l                The relocation of the drain line has no affect on the operation of the chemical radwaste system.
l                Therefore, this change will not increase the consequences or probability of a malfunction of
!-              " equipment important to safety previously evaluated in the USAR. An accident that involves the release of radioactive liquid in the Radwaste Building is described in the USAR Section 15.7.3, but this accident actually involves the release of radioactive gas from the spilled liquid. The L              L causes of such an accident are a leak in the tanks or operator error, not due to drain line location chenges." Therefore, the probability of occurrence or the consequences of any accident previously evaluated do not increase. The drain lines will function as before this change; thus, the possibility of creating an accident or malfunction not previously evaluated will not increase This activity -
does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditioris for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
USAR DESCRIPTION CORRECTION FOR FP TO WS INTERCONNECTION Log Number: 97-214
           ,     . Document Evaluated: USAR Section 9.5.1.2.2.2, Appendix E                                              )
3.1.2.6, & Figure 9.5.1 Sheet 8 US.AR Sections 9.5.1.2.2.2 and Appendix E 3.1.2.6.are being changed to reflect the correct             ;
description of the valve configuration between the Fire Protection -(FP) system and the Service       j Water (WS) system. The change will now state that the connection between the two systems               j contains a normally locked shut gate valve. The symbol for this valve (IFP035) is also being          i changed to reflect the normally shut position on USAR Figure 9.5-1, Si:cet 8. To eliminate any possible confusionin regards to FP system operating pressure, the phrase which provides the WS pump pressure ratings is being deleted from these paragraphs in the US AR. This change is to reflect the design shange that was made in 1986, at which time a jockey pump was added to             ;
maintain pressure in the FP system. Original design had WS maintaining standby pressure in FP,        :
and 1FP035 locked open. Thejockey pump was installed, because adequate pressure was not being maintained with other WS loads in operation. A full 10CFR50.59 Safety Evaluation was            j

o                                                             ,
1 Attaeme 2                 ,
to D W3276 r
Page 56 of 252                 ,
performed to make thejockey pump part of the permanent FP system design and close IFP035.
1 At that time, the Updated Safety Analysis Report (USAR) was changed to describe the addition                 i of thejockey pump, however, the description of the WS interconnection was not clarified to reflect the closed gate' valve.1 Valve IFP035 is opened only when the primary. fire pumps are impaired.LThe opening of IFP035 is controlled by Clinton Power Station (CPS) Procedure
                      ' 3213.01, " Fire Detection and Protection." The valve lineup table, CPS 3213.01V001, is being             -
changed to show iFP035 to be normally locked shut instead ofjust shut, however, the steps in this procedure are not being changed.
USAR Appendix F, Safe Shutdown Analysis, describes the FP system's function to protect safety-related equipment and the consequences of fire affecting the equipment. All design basis requirements are still met, and in fact, this change enhances this system's capability. Due to that fj fact, the probability of a rnalfunction er consequences of a malfunction of equipment important to           l safety is not increased. USAR Appendix F discusses postulated accident scenarios associated                  2 with fire, and takes into account the suppression capability of the FP systems. However, this change does not alter the suppression capability of the FP system.! Appendix F also evaluates the consequences in fire zones throughout the plant. This charge does not decrease the FP system suppression capability; therefore, the probability or the consequences of an accident previously            i evaluated in the USAR are not increased. As stated before, the capability of the FP water system
: does not diminish, no new equipment is being introduced, and the closed function of valve
                      .lFP035 ensures proper independent operation of both FP and WS. Therefore, there is no possibility of an accident or malfunction of equipment important t'o safety of a different type than previously evaluated in the USAR. The Technical Specifications specifically exclude requirements               .
            ^ for operation of the plant Fire Protection system, and states that these requirements will be                         l controlled by administrative procedures. This change does not diminish the protection of safety l
related structures, s'y stems, and components. Therefore, this activity does not reduce a margin of '
                                 ~
safety as defined in the basis for any Technical Specification.
DIVISION 11 125 VDC BATTERY CHARGER LOAD TEST Document Evaluated: PAC 0414-97 to CFS Procedure 9382.09                      Log Number: 97-215
  .                    Procedure Advance Change (PAC) 0414-97 made changes to Clinton Power Station (CPS)
                   - procedure 9382.09," Division II 125 VDC Battery Charger Load Test." PAC 0414-97 changed steps to allow staning of the battery charger using an external 125 VDC power supply; changed the load current to 310 amps DC to account for accuracy of BCT-2000 test equipment; added a step to take manual control of the test equipment to ensure that test current is maintained above 305 amps DC; and chang'ed the time requirement from 240 minutes to 260 minutes at 2 300 amps for the test equipment.to beloaded.
The external 125 VDC power supply is only temporarily installed to provide reference voltage to
                   - allow the battery charger to start. . No other operable or available equipment is affected by the
          ,            installation of the external 125 VDC power supply. This battery charger is designed to supply
                      ' load currents up to 375 amps at 130 VDC without damaging the charger, and it is designed to

1
  ,                                                                                   Attachment 2 to U-603276 Page 57 of 252 limit current to prevent damage. Normal operation of the DC system places the battery charger in       )
a condition of greater than 310 amps during an equalize charge. Increasing the time allows for the test equipment to stabilize at operating temperature. Operating the battery charger for longer amounts of time at the higher current is not outside of the design specifications of the equipment.
Therefore, this change will not increase the probability or consequences of a malfunction of           I equipment or an accident previously evaluated in th.e Updated Safety Analysis Report (USAR).
The battery charger is inoperable and isolated from any operational system by one AC breaker on the battery charger and an AC breaker on the AC feed Motor Control Center. The external 125 VDC power supply is removed before the equipment is restored for service. Therefore, these changes do not create the possibility of an accident of a different type than any evaluated previously in the USAR.                                                                                )
l The use of the external 125 VDC power supply only provides a method of bringing the equipment to a status for which the routine surveillance can be performed to satisfy Technical Specification requirements. The methods for satisfying Technical Specification requirements are not changed          j by the use of the external 125 VDC power. Technical Specifications require that the battery            I charger be able to supply a miriimum of 300 amps at a minimum of 125 VDC. By increasing the test current to 310 amps,it is ensured that Technical Specification requirements are satisfied.        {
This battery charger is designed to supply load currents up to 375 amps at 130 VDC without            !
damaging the charger for an indefinite amount of time. As a result, this activity does not reduce a   I margin of safety as defined in the basis for any Technical Specification.
DIVISION ll 125 VDC BATTERY CHARGER LOAD TEST l
Document Evaluated: PAC 0415-97 to CPS Procedure 9382.09C001                   Log No.: 97-216 i      Procedure Advance Change (PAC) 0415-97 made changes to CPS 9382.09C001," Division 11                  ;
125 VDC Battery Charger Load Test Checklist" PAC 0415-97 changed steps to allow starting of the battery charger using an external 125 VDC power supply and a step was added to take manual control of the test equipment to ensure that test current is maintained above 305 amps DC.
The external 125 VDC power supply is only temporarily installed to provide reference voltage to allow the battery charger to start. No other operable or available equipment is affected by the installation of the external 125 VDC power supply. This battery charger is designed to supply load currents up to 375 amps at 130 VDC without damaging the charger. Taking manual control of the test load current is only a change to the method of performing the test. Therefore, this change will not increase the probability or consequences of a malfunction of equipment or an accident previously evaluated in the Updated Safety Analysis Report (USAR). The battery charger is inoperable and isolated from any operational system by one AC breaker on the battery charger and an AC breaker on the AC feed Motor Control Center. The external 125 VDC power supply is removed before the equipment is restored for service. Therefore, th:se changes do not create the possibility of an accident of a different type than any evaluated previously in the US AR.
1 I

Attachment 2
* to U-603276 Page 58 of 252 The use of the external 125 VDC power supply only provides a method of bringing the equipment to a status for which the routine surveillance can be performed to satisfy Technical Specification requirements. The methods for satisfying Technical Specification requirements are not changed by the use of the external 125 VDC power. Taking manual control of the test load current is only a change to the method of performing the test. As a result, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
TS BASES FOR DIESEL GENERATOR SURVEILLANCE REQUIREMENTS Document Evaluated: TS Bases Change BL-97-015                                                                                                   Log Number: 97-217 Clarification is being added to the Technical Specification (TS) Bases for the Diesel Generator (DG) unloaded rapid start Surveillance Requirements (SR) regaramg the intent of achieving the specified voltage and frequency ranges within the 12-second time requirement during performance of the surveillance. The minimum voltage and frequency are achieved within the required 12 seconds when testing the DGs per SR 3.8.1.7, SR 3.8.1.12, SR 3.8.1.15, and SR 3.8.1.20.
However, the DGs are unloaded during these surveillance tests and when the DGs are rapid started without loading there is some overshoot and undershoot of voltage and frequency from the voltage and frequency ranges of the SRs caused by control system response characteristics.
According to the aesign of the DG controls, voltage and frequency will be maintained when loaded, and therefore, the performance is tested by those SRs that require loading. This change is inclunad as part of the corrective action for Condition Report 1-97-08-247 Four seatences are being added to provide clarification of the wording in the TS Bases. The change does not alter the design, function, or method of performing a function of a structure, system or component described in the Updated Safety Analysis Report (USAR). Therefore, the probability and the consequences of a malfunction of equipment imponant to safety evaluated previously in the USAR does not increase. Loss of the emergency diesel generators is an event evaluated in the USAR that might be considered likely to be affected by the change. However, this subject change does not affect the design criteria of the DGs The emergency diesel generators interface with systems that contain radioactive material, failures of these systems that could result in a release of radioactive material are postulated in the USAR. Since the change is only a minor clarification of the wording in the TS Bases and no other changes are being made that would increase the magnitude of the failures of these interfacing systems, it is concluded that the probability or consequences of an accident previously evaluated in the USAR does not increase. As a result of this change, no design or operational limits are challenged or exceeded, no equipment will be operated difTerently, and no new equipment is being installed. Thus, the change does not create the possibility of an accident or of a malfunction of equipment important to safety of a different type than previously evaluated in the USAR. Implementation of the change does not impact the fuel cladding, reactor coolant system, or primary containment-therefore, this activity does not reduce a margin of safety as defined in the basis for any To.
Specification.

  *                                                           .                       Attachment 2 to U-603276 Page 59 of 252 i     LEAVE RADWASTE BUILDING SUPPLY HEADER ISOLATION VALVE OPEN DURING NORMAL OPERATIONS 1
Document Evaluated: USAR Figure 9.2-12, Sheet 2                                 Log Number: 97-218 TG activity will change Updated Safety Analysis Report (USAR) Figure 9.2-12, Sheet 2 to show the Turbine Building Closed Cooling Water System (WT) Radwaste Building Supply Header Isolation valve (lWT067) as a normally open vah;e. WT is the normal supply to the condenser vacuum pumps seal water heat exchangers and the Component Cooling Water (CCW) system is an additional source. When the condenser vacuum pumps are not operating, Clinton Power Station (CPS) operates with 1WT067 open to minimize restoration time following a plant trip.
Maintaining IWT067 normally open oisures (Lat the WT header in the Radwaste Building remains full and will not require filling and venting following an event which requires a condenser vacuum pump to be started. When the condenser vacuum pumps are not in service, Radwaste Building Return Header Isolation valve (lWT066), Radwaste Building Return Header Isolation valve (lWT074), and CCW/WT Return Cross-connect valve (1WT075) will be closed, therefore, no flow will be allowed through the Radwaste WT header and operating with the 1WT067 valve open will have no impact on the WT system. Both the IWT067 and IWT066 are open during condenser vacuum pump operation and will not be impacted by this change. The condenser vacuum pumps use seal water to trap and remove air from the pump and the seal water heat exchangers prevent the seal water from boiling. This change will not impact the function or purpose of the seal water heat exchangers or operation of the condenser vacuum pumps.
This change will not cause the CCW system or the condenser vacuum pumps to be operated outside of their design or tested limits. The WT piping will be filled at all times, therefore, the connection between the WT and CCW systems will provide increased reliability and will reduce the time required to start the condenser vacuum pumps. Consequently, this change will not increase the probability or consequences of a malfunction of equipment important to safety evaluated previously in the USAR. When the WT system is configured with valve 1WT067 open and valve IWT066 closed, the condenser vacuum pumps are not in use. When the CCW system is used to cool the condenser vacuum purnps, valve IWT067 is closed. The probability of an accident is not increased, because the change does not impact the purpose or function of the WT     l or CCW systems, or the condenser vacuum pumps. With regard to improving the time required to start the condenser vacuum pumps, the consequences of an accident should decrease rat %
than increase when the pumps are used to mitigate an accident. The control of valve IWi a67 in CPS procedure 3204.01," Turbine Building Closed Cooling Water", will prevent auy different type of accident from occurring. The function or purpose of the WT or CCW systems, or the condenser vacuum pumps are not altered with this change. This change does not impact the function, purpose, or operation of the seal water heat exchangers and the failure of the heat exchangers are not discussed in the USAR. Therefore, this change will not create the possibility of an accident or a malfunction of equipment of a different type than prcviously evaluated in the USAR. The Technical Specification or Technical Specification Bases do not address the WT or CCW systems, or the condenser vacuum pumps and these systems have no nuclear safety related

U         <
Attachment 2           ,
to U-603276 Page 60 of 252 e   ' functions. Therefore, this activity does not reduce a margin of safety as defined in the basis for
    . any Technical Specification.;
DIVISION I 125 VDC BATTERY CHARGER LOAD TEST
    ! Document Evaluated: CPS No. 9382.08 &                                          Log Number: 97-220
                               ' CPS No. 9382.08C001 This chani;e involves several changes to Clinton Power Station procedure 93'82.0~8, " Division I
_.125 VDC Battery Charger Load Test." This change adds steps to allow starting ofthe battery charger using an ext::rnal 125 VDC power supply. .To account for accuracy of BCT-2000 test 1 equipment, theload current was changed to 310 amps DC. This change also includes a step to take manual control of the test equipment to ensure that test current is maintained above 305
    . amps DCc It changes the time requirement from 240 minutes to 260 minutes at 2 300 amps for the test equipment to be loaded.
    . The external 125 VDC power supply is only temporarily installed to provide reference voltage to allow the battery charger to start. No other operable or available equipment is affected by the installation of the external 125 VDC power supply. This battery charger is designed to supply load currents up to 375 amps at 130 VDC without daimaging the charger, and it is designed to
    ; limit current to prevent damage. Normal operation of the DC system places the battery charger in a condition of greater than 310 amps during an equalize charge. Increasing the time allows for
: the test equipment to stabilize at operating temperature. Operating the battery charger for longer amounts of time at the higher current is not outside of the design specifications of the equipment.
: Therefore, this change will not increase the probability or consequences of a malfunction of .
                                                                               ~
    ; equipment or an accident previously evaluated in the Updated Safety Analysis Report (USAR).
The battery charger is inoperable and isolated from any operational system by one AC breaker on
    . the battery charger end an AC breaker on the AC feed Motor Control Center. The external 125             ,
VDC power supplyis removed before the equipment is restored for service, and therefore, does            j
     .not create the possibility of an accident of a different type than any evaluated previously in the      {
     .USAR.-
The use of the external '125 VDC power supply only provides a method of bringing th'e equipment to a status for which the routine surveillance can be' performed to satisf9 Technical Specification requirements. The methods for satisfying Technical Specification rwements are not changed by the use of the external 125 VDC power. Technical Specifications require that the battery l charger be able to supply a minimum of 300 amps at a minimum of 125 VDC. By increasing the test current 10 310 amps, it is ensured that Technical Specification requirementr are satisfied.
This battery charger is designed to supply load currents up to 375 amps at 130 VDC without damaging the charger for an indefinite amount of time. As a result, this activity does not reduce a i margin 'of safety as definea in the basis for any Technical Specification.
l

  .                                                                                     Attachment 2 to U-603276 Page 61 of 252 DIVISION IV 125 VDC BATTERY CHARGER LOAD TEST                                                        l Document Evaluated: CPS No. 9382.11 &                                           Log Number: 97-221 CPS No. 9382.11C00I                                                       i
                                                                                                             )
This proposed change involves several changes to Clinton Power Station procedure 9382.11, .
        " Division IV 125 VDC Battery Charger Load Test." This change adds steps to allow starting of the battery charger using an external 125 VDC power supply. To account for accuracy of BCT-2000 test equipment, the load current was changed to 310 amps DC. This change also               l includes a step to take manual control of the test equipment to ensure that test current is maintained above 305 amps DC. It changes the time requirement from 240 minutes to 260 minutes at 2 300 amps for the test equipment to be loaded.
The external 125 VDC power supply is only temporarily installed to provide reference voltage to allow the battery charger to start. No other operable or available equipment is affected by the installation of the external 125 VDC power supply. This battery charger is designed to supply load currents up to 375 amps at 130 VDC without damaging the charger, and it is designed to limit current to prevent damage. Normal operation of the DC system places the battery charger i~n a condition of greater than 310 amps during an equalize charge. Increasing the time allows for the test equipment to stabilize at operating temperature. Operating the battery charger for longer amounts of time at the higher current is not outside of the design specifications of the equipment.
l Therefore, this change will not increase the probability or consequences of a malfunction of equipment or an accident previously evaluated in the Updated Safety Analysis Report (USAR).
The battery charger is inoperable and isolated from any operational system by one AC breaker on l        the battery charger and an AC breaker on the AC feed Motor Control Center. The external 125 l        VDC power supply is removed before the equipment is restored for service, and therefore, does i        not create the possibility of an accident of a different type than any evaluated previously in the USAR.
The use of the external 125 VDC power supply only provides a method of bringing the equipment to a status for which the routine surveillance can be performed to satisfy Technical Specification requirements. The methods for satisfying Technical Specification requirements are not changed by the use of the external 125 VDC power. Technical Specifications require that the battery charger be able to supply a minimum of 300 amps at a minimum of 125 VDC. By increasing the test current to 310 amps, it is ensured that Technical Specification requirements are satisfied.
This battery charger is designed to supply load currents up to 375 amps at 130 VDC without damaging the charger for an indefinite amount of time. As a result, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.

,   [                     >                                                              Attachment 2          * )
to U-603276             I Page 62 of 252 J
ADDITION O'F AN ASSISTANT MANAGER - CPS POSITION Document Evaluated: USAR Change 8-069                                            Log Number: 97-222 This change incorporates a new position of Assistant Manager - Clinton Power Station (CPS)in the plant'stafforganization.: Updated Safety Analysis Report (USAR) Section 13.1.2.2.1, Table
:13.1-1, and Figure 13.1-3 are impacted. The new position provides the Manager -CPS more
      - opportunities to interact with direct reports, and to focus on operational concerns. Creation of a new position to assist in the administration of activities at CPS does not negatively impact the
      . operation or maintenance of equipment important to safety. No change to existmg programs or responsibilities of the Manager - CPS are being implemented by this change. Since there is no
      ' identified impact on plant operations, this change does not increase the probability or consequences of an accident previously evaluated in the US AR. This administrative change does not impact the possibility of an accident or a malfunction of equipment important to safety of a different type, because there are no changes to existing programs or responsibilities. This activity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation or structures, systems, or components subject to these requirements. Therefore, this activity does not reduce'a margin of safety as defined in the basis for any Technical Specification.
TEMPORARY MODIFICATION FOR GATEHOUSE MCC "A" (0AP32E)
_ Document Evaluated: Temporary Modification 97-091                               Log Number: 97-223-Unit Sub E (OAP45E) Cubicle 6B is the normal power supply to Gatehouse Motor Control Center (GH MCC)"A" (0AP32E). - During a Unit Sub E bus outage, GH MCC A requires a temporaiy feed from the 12KV loop in' order to stay in service. Temporary Modification (TM) 97-091 affects the configuration of Updated Safety Analysis Report (USAR) Figure 8.3-1. This TM does not impact the operation of the 12KV loop or the' Auxiliary Power system.
GH MCC A, Unit Sub E and the 12KV loop are non-safety related and only impact safety related systems through association with the distribution system via design basis calculations. Providing power to the MCC from the 12KV loop reduces plant loading from the Reserve Auxiliary Transformer (RAT), and thus, is conservative with respect to plant loads. Therefore, the proposed change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The decreased loading on the electrical
      - distribution system does not increase the probability and consequences of accidents previously
: evaluated in the USAR. . Connection of a non-safety related MCC to a different non-safety related
      ' source does not create.the possibility c f creating an accident or malfunction of equipment
      , important to safety of a different type. This activity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation or the components subject to these requirements. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Techcical Specification.
4

c                                                                                         Attachment 2 to U-603276 Page 63 of 252
                                                      .       .                                                         .)
TEMPORARY MODIFICATION FOR GATEHOUSE MCC "B" (OAP33E)
Document Evaluated: Temporary Modification'97-092                                Log Number: 97-224 Unit Sub F (OAP46E) Cubicle.4A is the norma: power supply to Gatehouse Motor Control Center
: (GH MCC)"B" (0AP33E)~ During a Unit Sub F bus outage, GH MCC B requires a temporary feed from the 12KV loop in order to stay in service. Temporar Modification (TM) 97-092 affects'the configuration of Updated Safety Analysis Report (USAR) Figure 8.3-1. This TM does J not impact the operation of the 12KV loop or the Auxiliary Power system.
GH .MCC B, Unit Sub F and the 12KV loop are non-safety related and only impac safety related              i
             - systems through association with the distribution system via design basis calculations. Providing power to the MCC from the 12KV loop reduces plant loading from the Reserve Auxiliary                      i J
             . Transformer (RAT), and thus, is conservative with respect to plant loads. Therefore, the -
proposed change does not increase the probability or consequences of a malfunction of equipment
!              important to safety previously evaluated in the USAR.' The decreased loading on the electrical            j
             . distribution system does not increase the probability and consequences of accidents previously            I evaluated in the USARJ Connection of a non-safety related MCC to a different non-safety related source does not create the possibility of creating an' accident or malfunction of equipment L               important to safety of a different type. This activity does not impact any Technical Specification
(   ,           safety limits, limiting safety system settings, or limiting conditions for operation or the L               components . subject to these requirements. Therefore, this activity does not reduce a margin of
:             safety as defined in the basis for any Technical Specification.
BATTERY AND DC DISTRIBUTION TEMPORARY MODIFICATION Documents Evaluated:' PACS 0451-97 & 0453-97                                    Log Number: 97-225
              . Procedure Advance Changes (PAC) 0451-97 and 0453-97 allow the installation of a temporary modification which cross-ties the IF battery charger to the Division I DC bus. This temporary modification maintains a float charge on the Division I battery while the Division I battery charger is out of service for maintenance. .While this temporary modification is in place, the Division I DC distribution system will be declared inoperable.
The cross-tie provides an alternate source of power to the 1 A DC Motor Control Center (MCC) within the design rating cf the 1 A bus. The alternate source of pr wer does not connect to any other safety related source. Although the IF battery charger w;t provide power within the design ratings of the l' A bus, a failure of the IF charger could re.cQt m an output voltage that exceeds the maximum voltage rating of the 1 A bus. Although the Pavision I and IF chargers are equipped
: with similar high voltage shutdoivn protective cirants, the non-Class IE charger protection circuit would be relied u' pon to provide this protecti,o feature for the Class IE bus. However, since the Division I DC distribution' system would'oe considered inoperable, this equipment would not be credited as important to safety as evaluated in the Updated Safety Analysis Report (USAR).
Therefore, this ch .nge %c not increase the probability or consequences of a malfunction of

~
                                                                                                       . l Attachment 2 to U-603276 Page 64 of 252 equipment important to safety as previously evaluated in the USAR. Since all equipment will be operated within design limits and the temporary modification meets the requirements ofisolating the Division I bus in conformance with the USAR, this change does not increase the probability or consequences of an accident previously evaluated. Per Technical Specification 3.8.5, this temporary modification will only be in place during Modes 4 and 5 with no irradiated fuel movement. Also, per Technical Specification 3.8.5, the Division I DC distribution system will be declared inoperable. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
OFFSITE DOSE CALCULATION MANUAL (ODCM) CHANGE                                                           ,
Document Evaluated: ODCM                                                       Log Number: 97-226 This change implements revisions to Table 3.9.2-2 and Table 3.9.2-2 Table Notations in the Offsite Dose Calculation Manual (ODCM). This revision adds item 3 to the table for Inservice Calibrations on Stati0n Heating, Ventilation, and Air-Conditioning Exhaust and Standby Gas Treatment System Exhaust Process Range Monitors (PRMs). Table Notations "c" and "d" add an annual channel calibration requirement and applicability. The changes implement Regulatory Guide 1.21, Section C.5 requirements for calibrating continuous monitors using comparisons               I against analysis of specific radionuclides.
These changes relate to surveillance and monitoring of radioactive efiluents and calculations of dose to the public, not to operation of plant equipment. Therefore, this c' ,e does not impact the probability or consequences of a malfunction of equipment importat .o safety previously evaluated in the Updated Safety Analysis Report (USAR). The USAR does not evaluate
  - accidents associated with radioactive effluent monitor comparisons; therefore, the probability of an accident does not increase. The USAR evaluates the consequences of accidents associated with dose to unrestricted areas as a result of radioactive effluents. This revision improves dose determinations from radioactive gaseous efiluents by ensuring 'the PRMs are calibrated with grab samples. Since the ODCM determines the dose related to radioactive efIluents from both normal plant operation and during postulated accidents, the ODCM itselfis not capable ofincreasing the consequences of an accident. There are no credible accidents or failure modes for equipment important to safety associated with these changes to the ODCM. These changes to the ODCM                  i bring the ODCM in compliance with Regulatory Guide 1.21, Section C.5 requirements and Technical Specifications. Therefore, this activity does not reduce a margin of safety as defined in       j the basis for any Technical Specification.                                                               j i
ll l
l

Attachment 2 to U-603276 Page 65 of 252 DISCONTINUING SOME WATER CHEMISTRY MEASUREMENTS FOR CONDENSATE AND FEEDWATER Document Evaluated: USAR Change 8-070                                         Log Number: 97-228 (ref. ECN 28678)
Engineering Change Notice (ECN) 28678 removed turbidity cells l AE-PS026/029 and silica analyzers I AIT-PS024/025 from panel IPL88JA. Changes were made to Updated Safety Analysis Report (USAR) Figure 9.3-4 in accorda r with ECN 28678. This change removes the footnotes at the end of the design bases for the condensate cleanup system output in USAR section 10.4.6.1, Measurements are not typically taken for silica, chloride, and sodium.
However, the footnotes state that certain parameters are not measured when the conductivity is within certain limits. This change removes the silica analyzers and the turbidity cells and performing measurements implied by section 10.4.6.1 for the condensate and feedwater systems.
Assuming that not monitoring the chemistry parameters being deleted by this change could allow detrimental water in the loop, the chemistry measurements taken for the reactor water would detect a problem and allow for coaective action. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety evaluated               ,
previously in the USAR or of a different type. Poor water chemistry could cause corrosion or          j plating of deposits on components. The condensate and feedwater systems are not accident               l
                                                                                                          ]
mitigating systems; and the water chemistry for accident mitigating systems is monitored by the measurements taken for the reactor water. Therefore, this change does not increase the l
probability or consequences of an accident previously evaluated in the USAR or of a different type. Conductivity measurements or acceptance standards are not changed by this revision.
!   Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical l   Specification.
1 PERMANENT SLAB IN PROTECTED AREA                                                                      I l
Document Evaluated: ECN 30604                                                 Log Number: 97-230 Engineering Change Notice 30604 installs a permanent reinforced concrete slab adjacent to and east of the Plant Support Building. The Plant Support Building is within the Protected Area about 300' south of the Fuel Building. The installation of this slab impacts Updated Safety           l t    Analysis Report (USAR) Figures 1.2-3, Sheet I and 2.1-7. The concrete slab will have no interface with any plant equipment, therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the L 3AR. Chapter 3 of the USAR identifies external events that may initiate an equipment failure.
The concrete slab is approximately 300 feet from any safety related structures. It consists of 10 monolithic pours averaging 36' X 30' and weighing between 80-100,00 pounds. The slab can not collapse, fall, slide or be blown into any permanent plant equipment. During a seismic event, minor movement between the pours or cracking may occur, however, it will not effect any plant equipment. Therefore, this change does not increase the probability or consequences of an

4 Attachment 3 to U-603276 Page 66 of 252
                  - accident previously evaluated in the USAR.;No new types of accidents or malfunctions are created, because the slab is immobile and passive. This activity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation.;
Therefore,1this activi'y does not reduce a margin of safety as dermed in the basis for any Technical
                 . Specification;
                 ~ REACTOR WATER CLEANUP FILTER DEMINERALIZER EFFLUENT CONDUCTIVITY
: METER REMOVAL Document Evaluated: 'JCN 30303                                                   Log Number: 97-231
                 - Engineering Change Notice (ECN) 30303 implements Modification PSF 02'8, Supplement 1,
              >      which removes the Reactor Water Cleanup (RWCU) Filter /Demineralizer (F/D) effluent conductivity meters from local panel 1G36P002. This ECN is a result of circuit loading problems
                 . due to'the input impedance of the General Electric (GE) 180 meters. When combined with the dropping resistor, recorder, trip units, and computer input; the low impedance causes the circuit-
: equivalent resistance to be too low to allow it to be calibrated within loop tolerances. The increase in resistance resulting from removing the GE 180 meters allows an acceptable loop calibration. The removal of these meters impacts Updated Safbty Analysis Report (USAR)
: Figure 5.4-18.:        ,
                 . Section 5 of the Operational Requirements Manual refers to the reactor water conductivity .
meters, however, the removal of the local panel conductivity meters does not affect the reactor water conductivity measurements. Therefore, this change does not increase the probability or
: consequences of a malfunction of equipment important to safety previously evaluated in the -
:    USAR. RWCU F/Ds effluent conductivity instrument loops are affected by the removal of the GE 180 meters. These conductivity loops determine individual F/D resin performance and do not affect the safety related reactor water conductivity instrument loops. The conductivity of the F/D will be displayed on a panel in Containment instead of on the local panel. This change does not
                  ' affect any other alarms, chart recorders or computer points; and therefore, the lack oflocal F/D conductivity does not affect the probability or consequences of an accident previously evaluated
                  . in the USAR. Because these meters do not have interlocks associated with them, are passive in -
nature, and are not relied upon for the operation of any safety related systems, this activity does
: not create the po_ssibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. RWCU F/D conductivity or Process Sampling systems ability to measure the F/D conductivities arr; not discussed in the Technical Specifications; therefore, this activity does not reduce a margin of safety as defined in the basis for
                . any Technical Specification.
                           'a e

1
     .                                                                                         Attachment 2
{
to U-603276 Page 67 of 252 CIRCULATING WATER PUMP TEMPORARY SHELTER                                                                l Document Evaluated: Temporary Modification 97-06 Log Number: 97-232 Temporary Modification 97-06 provides a shelter for the Circulating Water (CW) pumps and motors to keep off snow and ice. The shelter has temporary e!cctrically powered heaters to              q prevent' water bearing cooling lines from freezing. The metal, non-flammable building is 18 feet        I wide by 12 feet high by 57 feet long. Plant Security requires temporary lighting outside the
           - building The modification is temporary and is only for use while the plant is in Modes 4 or 5 If the shelter fails, a trip or failure of the CW pumps could occur. However, even though the CW pumps would cause the plant to trip, the temporary modification will only be in use through Modes 4 or 5. In these Modes, the structure can not initiate an evaluated accident. Updated Safety Analysis Report (USAR) Section 10.4.5.1.1 states that the CW pumps are not required for         ;
nuclear safety. Therefore, this change does not increase the probability or consequences of a malfunction of equipmem important to safety previously evaluated in the USAR. The most likely I
accident that could occur is the shelter collapsing under a wind and/or snow load in turn causing the CW pumps to trip. However, as stated before, this Safety Evaluation only covers the temporary modification while the plant is in Modes 4 or 5. Therefore, this temporary modification does not increase the probability or consequences of an accident previously evaluated in the USAR, Nor, is there a credible accident scenario or malfimetion of equipment of I
a different type than previously evaluated. This activity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation.
Therefore. this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
l USAR CHANGES TO REFLECT NEW MATERIAL FOR SPLICES AND CABLE I            TERMINATIONS                                                                                           l Document Evaluated: ECN 30448 and USAR Change 8-086                            Log Number: 97-233 R/l Engineering Change Notice (ECN) 30448 allows the use of a new silicon rubber tape for                 !
insulating cable terminations and splices in non-harsh zones of the plant for safety and non-safety related equipment. The qualification testing for this material was done by the vendor and does l             not agree with the Updated Safety Analysis Report (USAR) conformance statements for Regulatory Guide 1.131. This change expands the conformance statements in Section 8.1.6.1.24 of the USAR. The tape authorized by this ECN is only used in locations where analysis shows that it can perform its design function. The limitations upon use of this material are provided to the installers in Clinton Power Station (CPS) procedure 8493.01, " Terminations With UCI Splicing Tape." Since the tape is authorized for use at CPS on the electrical terminations of any plant system, any system important to safety that has electrical terminations could potentially be    :
I affected. Properly installed,'the tape will not fail during its qualified life and will not create an electrical short or other failure of the equipment connected to the taped splice or termination.
Therefore, this change does not increase the probability or consequences of an accident or l

Attachment 2 to U-603276
~
Page 68 of 252.
equipment malfunction previously evaluated in the USAR. The new tape performs the same function as the previously referenced Okonite and the Raychem taping material. When utilized in locations where analysis shows that it can perform its design function, this tape performs in the -
: same manner as the material previously' authorized. The new material does not introduce any new failure modes for electrical terminations. Therefore, this change does not create the possibility of L an accident or equipment malfunction of a different type. The Technical Specifications do not address the manner in which electrical terminations are made but does require that equipment be operable l When used in accordance with the requirements established through the tape's qualification repon, the equipment will be operable. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
USAR CHANGE FOR THE MAINTENANCE IMPROVEMENT GROUP LEADER' Document Evaluated: USAR Change 8-061                                         Log Number: 97-234
                    ' ' This change to the Updated Safety Analysis Report (USAR) adds the Maintenance Improvement 1
Group Leader d.ne to the formation of the Maintenance Improvement Group. This change Aremoves the Fix It-Now'(FIN) Team from the direct supervision of the Supervisor of.
                      . Maintenance Phmning, and also, removes reference of the FIN Team from the USAR. U'SAR
: Section 13.1.2. 2.4 and Figure 13.1-3 corresponds to this change. The formation of the Maintenance Improvement Group and Maintenance Improvement Group Leader, and deletion of s the FIN Team from the USAR is an administrative change. Activities of the Maintenance Improvement Group will continue the same as when under the direction of the Supervisor of
: Maintenance Planning, only in more depth: Existing administrative procedures cover activities of L                        the FIN Teami Since these administrative changes are not directly related to equipment '
                      ' manipulation, there is no impact on the probability or consequences of a malfunction of equipment important to st.fety previously evaluated in the USAR. Activities of the Maintenance .
                       ; Improvement Group are performed to ensure that accidents will not occur and the FIN Team responsibilities hav'e not changed. Therefore, this change does not increase the probability or consequences of a previously evaluated or new accident from occurring. The Technical Specifications do not address the activities of the Maintenance Improvement Group or the FIN Teara; Technical Specification requirements for stafling and qualification are not affected by this
                       , change. :Therefore, this activity does not reduce a margin of safety as defined in the basis for any i              Technical Specification.
NUCLEAR REVIEW AND AUDIT GROUP CHARTER Document Evaluatedi Procedure NRAG 101-1, .                                   Log Number: 97-235
                                                  ' USAR Change 8-083, and .
                                     -               ORM Change 17-2
                                                                ~
z This procedur~e revision clarifies an administrative function for the Nuclear Review and Audit Group (NRAG). - This change revises NRAGP 101-1, Paragraph 6.1, Part i. from "Repons and meeting minutes of the Facility Review Gioup (FRG) and the Independent Safety Engineering o.
i l
r

[
: i.           ,,.
     -.                                                                                            Attachment 2 to U-603276 Page 69 of 252 i
Group (ISEG)" to " Meeting minutes of the Facility Review Group (FRG)" and "n. Reports of the Independent Safety Engineering Group (ISEG)",'- Updated Safety Analysis Report (USAR)
: j.                ' Paragraph 13.4.2, Part d a' nd Operations Requirements Manual, Paragraph 6.5.2.7, Pan i relate to l-               ' this change.L This change only clarifies the scope of review for the NRAG and does not impact or create equipment malfunctions. The change clarifies that the NRAG reviews FRG meeting minutes and ISEG reports 'which serve as an after the fact review, and therefore, has no possibility ofimpacting or creating an accident. The scope of the change is administrative review of a non-Technical Specification activity, and therefo're, does not reduce a margin of safety as defined L                   m the basis for any Technical Specification.
L l"                  BATTERY AND DC DISTRIBUTION TEMPORARY BATTERY CHARGER''
l f                   Document Evaluated: PAC 0492-97 for CPS Procedure.3503.01                      Log Number:,97-237 R
Procedure Advance Change (PAC) 0492-97 allows the installation of a temporary connection l-Jwhich cross-ties the IF battery charger to the Division IV DC bus. This maintains a float charge
_ on the Division !V battery while the Division IV battery charger is out of service for maintenance.
                  - While this temporary connection is in place, the Division IV DC distribution system will be declared inoperable.                                                                                 . 4 1
J
                  . The cross-tie provides an alternate source of power to the Division 4 Motor Control Center ID L                    within design ratings. ;The alternate source of power does not connect to any other safety related
                   ' source. Even though the IF battery charger will provide power within the design ratings of the -
l
                  , Division 4 DC bus, a failure of the IF charger could result in an. output voltage that exceeds the maximum voltage rating of the Division 4 DC bus. Although the Division 4 and IF chargers are l                    equipped with similar high voltage shutdown protective circuits, the non-Class lE charger l:                   protection circuit would be relied upon to provide this protective feature for the Class lE bus.
                  . However, since the Division 4 DC distribution system would be considered inoperable, this equipment would not be considered important to safety as evaluated in the Updated Safety
                   . Analysis Report (USAR). Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety as previously evaluated in the USAR. Since all' equipment will be operated within design limits and the temporary modification
                   ' meets the requirements ofisolating the Division 4 bus in conformance' with the USAR, this change L,                   does not'_ increase the probability or consequences of an accident previously evaluated. Per L                 . Technical Specification 3.8.5, this temp'orary connection will only be in place during Modes 4 and 5 with no irradiated fuel movement. Also, 'per Technical Specification 3.8.5, the Division 4 DC distribution system will be declared inoperable. Therefore, this activity does not reduce a margin L           T     ' of safety as defined in the basis for any Technical Specification.'      -
t i
L!

a Attachment 3 to U-603276.
Page 70 of 252 UPDATE USAR TO AGREE WITH DESIGN BASIS ROOM TEMPERATURES Document Evaluatedi ECN 30447; USAR Change 8-264                                  Log Number: 97-238 i Updated Safety Analysis Report (USAR) Table 3.11-5, Notes 18 and 20 do not specify the correct design basis shutdown' temperatures for Reactor Core Isolation Cooling (RCIC) and
: Emergency Core Cooling System (ECCS) pump and heat exchanger rooms. . Calculation VY-45
         - Revision 0 est'a blished the temperatures in the design criteria document (DC-ME-09-CP) for
         .- RCIC and ECCS pump and heat exchanger rooms. DC-ME-09-CP incorrectly references a -
l        1 revision of calculation EQ-10 that is no longer applicable. Engineering Change Notice (ECN)
          - 30447 revises DC-ME-09-CP to reflect the correct design basis references and temperature of heat exchanger room H-ll in accordance with the latest revision of VY-45. ECN 30447 also c revises USAR Table 3.11-5.
i This change is largely' editorial or an improvement in area temperatures. Any failure of plant equipment would have the same consequences." Therefore, this change does not increase the jrobability'or consequences of a malfunction of equipment important to safety previously .
          ' evaluated in the USAR. The temperatures in DC-ME-09-CP are the result of various operating situations including accidents. Potential affects of changes in area temperatures on Equipment Qualification have been reconciled in accordance with Engineering Instruction DE-22. If
         . equipment qualification is maintained, then the probability or consequences of an accident do not
         . increase. The equipment should operate as expected and there is not an increase in the probability (of creating 'an accident or equipment malfunction.- This activity'does not impact any Technical 1
          ;; Specification safety limits, limiting safety system settings, or limiting conditions for operation.
TTherefore, this activity does not reduce a margin of safety as defined in the basis for any Technical
          ' Specification. 4
         . QA PROGRAM REVISIONS
          - Document Evaluated: ' QAM Revision 26,                                            Log Number: 97-239 m                 USAR Change 8-073, and ORM Change 21-1 This Safety Evaluation makes four changes to licensing basis documents. Changes one and two complete the formation of the Nuclear Safety and Performance Improvement (NSPI) Department.
Change' one' changes the reporting for the Director-Licensing from the Vice President to the
          ~ Manager-NSPIJ Change two changes the reporting for ISEG from the Manager-Quality (Assurance (QA) to the Manager-NSPI. Change three eliminates references to "QA .surveillances" and replaces the term with "QA assessments". Change four combines the positions of Manager-Nuclear Support Services and Manager-Nuclear Training to form the Manager-Nuclear Training
: and Support, These changes are administrative changes that do not affect the design, materials, construction, maintenance,' operation, modification, or reliability of plant systems, stmetures, and components. The responsibilities, functions, staffing, or qualifications'of the personnel involved r   .do not change,just the reporting level or the title of the person to whom they report. Therefore,

  .                                                                                   Attachment 2 to IM03276 Page 71 of 252 the changes do not impact the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR or of a different type. For the same reasons, the changes do not impact any accidents analyses considered in Chapter 6 or 15 of the USAR, nor do they have a potential of creating a new type of accident. These changes are administrative changes to organizations and administrative controls that were not used u the basis for any Technical Specifications. Therefore, this activity does not reduce a margin of safety as    -
defined in the basis for any Technical Specification.
ADDITION OF REQUIREMENT TO OBTAIN NRC REVIEW AND APPROVAL PRIOR TO CHANGING THE REACTOR VESSEL SPECIMEN WITHDRAWAL SCHEDULE Document Evaluated: TS Bases Change BL-97-019                                   Log Number: 97-240 The purpose of this change is to add clarification to the Technical Specification Bases Section B3.4.11,"RCS Pressure and Temperature (P/T) Limits," and Updated Safety Analysis Report (USAR) Section 5.3.1.6.1. The clarification requires Nuclear Regulatory Commission (NRC) approval for any changes to the Reactor Vessel Specimen Withdrawal Schedule before implementation. Also, the NRC requires a license amendment for any changes to the specimen withdrawal schedule, which do not conform to the American Society for Testing Materials
       " Standard Practice for Conducting Surveillance Tests For Light Water Cooled Reactors" (ASTM E-185) standard. Finally, the clarification adds NRC Administrative Letter (AL) 97-04, "NRC Stafr Approval for Changes to 10 CFR Part 50, Appendix H, Reactor Vessel Surveillance Specimen Withdrawal Schedules," as reference 10 to the end of Bases section B3.4.11.
Section 5.3.2.2 of the USAR describes the most severe operating upset transient to which the vessel may be subjected. The most adverse temperature and pressure anywhere in the vessel head and/or shell areas occuts in the bottom head. A transient of this type would be terminated by a high pressure scram well below the P/T curve limits. The most limiting event for overpressure protection is the closure of all main steam line isolation valves with a secondary scram. The AL 97-04 guidance does not introduce any mechanism that can act as initiator for the pressure transients, change any systems or components in the reactor vessel, or cause operation outside design parameters. NRC must approve a postulated change to the specimen withdrawal schedule before implementation. The submittal and approval will be done prior to and separate from the refueling outage in which the specimen is withdrawn. Therefore, the proposed change does not        !
increase the probability or consequences of an accident previously evaluated in the USAR Title 10 Part 50, Appendix H, of the Code of Federal Regulations (CFR)," Materials Surveillance Program," monitors the reactor vessel for any changes in the fracture toughness properties of the ferritic materials used in construction. Requiring NRC approval prior to changing the specimen withdrawal s:hedule, which is specified in USAR Section 5.3.1.6.1, complies with 10 CFR 50,
!       Appendix H. Obtaining NRC approval prior to changing the schedule will not affect any seismic, environmental, quality or separation criteria of any structure, system or component important to safety. Therefore, this change does not increase the probability or consequences of a malfunction i       of equipment important to safety previously evaluated in the USAR. This change does not I
introduce or modify any SSCs or failure modes, and therefore, does not create the possibility of i
1                                                                                                           !
i i
l
                                                                                                            )

Q-
  ,                                                                                     , Attachment 2 to U-603276
               <     <                                                                    Page 72 of 252 n
an accident _or malfunction of equipment imponant to safety of a different type. The proposed addidon of the AL 97 04 guidance preserves the specimen sampling frequency accepted by the-L NRC. The change will not affect the current P/T limits, the determination of new limits, the protective functions that' terminate the transients de' scribed in the USAR, nor will it affect the requirements of the Limiting Condition for Operation of 3.4.11. Therefore, this activity does not
        - reduce a margin of safety as' defined in the basis for any Technical Specification.
USAR CHANGE TO OPERATING EXPERIENCE FEEDBACK PROGRAM
        ~ Document Evaluated: USAR Change 8-078                                           Log Numberi 97-241 This chans,e revises Updated Safety Analysis Repon (USAR) Appendix D, Item I.C.5,                     ,
          " Procedures for Feedback of Operating Experience to Plant Staff," This revision transfers responsibility for the operating experience feedback program from the Licensing Depanment to the Manager-Nuclear Safety and Performance Improvement (NSPI). Also,'this revision deletes a description of the operating experience feedback program in USAR Section 13.5.1.3 that existed prior to the first fuel loading. : The purpose.of the operating experience feedback program is to assure that events and fimd ngs at Clinton Power Station (CPS) and other utilities are evaluated for any corrective actions that should be taken at CPS. The change transfers the responsibility of this program from Licensing to the Manager of NSPI. Elements of tne program will be perforrr i under the Independent Safety Engineering Group (ISEG), the Licensing Department, and the Corrective Action Program. The Manager of NSPI will bring more resources to the program than were previously available, and members of the ISEG are as qualified as members of the Licensing
        . Department to evaluate operating experience. Therefore, the operating experience feedback' program should be strengthened by this transfer of responsibility and does not increase the probability of a malfunction of equipment important to safety. The' requirements for 50,59 Safety
        ' Evaluations of recommended plant changes remain the same, therefore, the revision does not increase the consequences of an equipment malfunction. The 'USAR requires that operating experience be included in plant operating practices and equipment improvements, this requirement assures that this revision does not increase the probability or consequences of an accident previously evaluated in the USAR.- One of the main purpose:, of the Operating Experience Feedback Program is to identify the possibility of accidents and equipment malfunctions different than those previously evaluated. By transferring the responsibility of this program to the Manager of NSPI, more resources are available to review operating experience. Therefore, this revision does not create the possibility of an accident or equipment malfunction of a different type than previously evaluated in the USAR. The Technical Specifications' do not address the Operating Experience Feedback Program, the administrative responsibilities of the Licensing Department or Independent Safety Engineering Group. Therefore, this activity does not reduce a margin of safety'as dermed in the basis of any Technical Specification.
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   .                                                                                  Attachment 2 l-                                                                                    to U-603276 l                                                                                     Page 73 of 252 l
APPENDIX R EVALUATION OF FIRE AREA D-8
!      Activity Evr.luated: Modification FP-101 and                                   Log Number: 97-242 l
USAR Change 8-080 nre Area D-8 consists of the Division 1 Diesel Generator (DG) ventilation fan room and air           -
intake located along the south wall of the DG building at elevation 762' However, the Division 2 DG power output cables are also routed through area D-8. These cables were previously considered in the Updated Safety Analysis Report (USAR) to be protected by a fire barrier              ;
(Thermo-Lag). Due to Nuclear Regulatory Commission and industry concerns over Thermo-Lag's ability to satisfy 10CFR50 Appendix R fire barrier requirements, Thermo-Lag can no longer be relied upon to adequately protect the cables in the event of a fire. As a result, both Safe Shutdown methods I and 2 would be unavailable in the event of a fire in area D-8 concurrent with a Loss of Offsite Power (LOOP). A coincident LOOP for all evaluated fires is currently a Clinton Power Station assumption in USAR, Appendix F, Section 1.4.B.4; however, this assumption is not required for all fires by regulatory requirements. This activity demonstrates that offsite power would be available to support safe shutdown utilizing Method 1 or 2 during an Appendix R fire in Fire Area D-8. This activity also reduces the combustible loading and the number of cables in this fire area by abandoning the breathing air compressors and associated components in place. Combustible items associated with this equipment will be removed.                l l
There are no failure modes associated with the portions of the breathing air system that are being abandoned / retired in ph ce that could impact any assumed accident initiators. The affected components will be isolated from the plant service water system and fire protection systern by closing the isolation valves to the affected components. The abandoned / retired in place breathing air components will not adversely impact any operable systems or components. The Thermo-Lag           i abandoned on the DG output cables will not affect DG operation or any accident initiators. The revised Appendix R analysis will not introduce any failure modes that could affect any previously analyzed accident initiators. The Breather Air Compressors, filters, deluge valve and associated      i components and instrumentation perform no safety functions during any transients or design basis accidents. The safety related Main Control Room Emergency Breathing Air Storage system and the Breathing Air system containment isolation valves are.not affected by this modification. The abandoned / retired in place equipment will be isolated from any operable systems by closing any interface system's isolation valves and by disconnecting and removing electrical power and control cables. The Thermo-Lag abandoned on the DG output cables will have no afTect on DG operation and will not prevent the diesel from performing its safety functions. The offsite power evaluation for Fire Area D-8 demonstrates that offsite power would be available to support safe shutdown utilizing Method 1 or 2 during an Appendix R fire in area D-8. The Appendix R safe shutdown systems and components powered from the 4160V 1 Al and IBl buses are not affected by this modification. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
The abandoned / retired in place Breathing Air Compressors, filters deluge valve, and associated components and instrumentation will be isolated from any operable systems or components, and combustible materials will be removed. The abandoned components will not introduce any new

Attachment 2 to U-603276 Page 74 of 252 credible failures or accident initiators. Abandoning the Thermo-Lag installation in place has no effect on the seismic qualification or ampacity rating of the cables and their supports. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important          !
to safety of a different type than previously evaluated. The equipment that is abandoned / retired in place will not adversely impact the ability of any safety related systems or components to mitigate the consequences of any accidents or transients. Nor .will it prevent the diesel from performing          -
i any ofits safety functions or affect DG operation._ Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
USAR CHANGES IN DRYWELL PURGING POWER GENERATION DESIGN BASIS DURING DIFFERENT MODE
  . Document Evaluated: USAR Change 8-081
_                                                      Log Number: 97-244           l The proposed change revises two sentences in Updated Safety Analysis Report (USAR) Section 9.4.7.2.1.2.a. The first sentence in this section will read: "Drywell purging for airborne activity control is not permissible during plant operating Modes 1,2, or 3 since drywell supply air penetration / isolation valves iVQ01 A and 1VQOlB must remain closed." This change brings the USAR into agreement with Drywell Technical Specification Bases B3.6.5.3, and Clinton Power Station (CPS) Procedure 3408.01, Containment /Drywell Heating, Ventilating and Air Conditioning (HVAC) Systems Operating. This change is a more appropriate discussion for not
_ performing containment /drywell purging in Modes I.,2, or 3. The last sentence in Section 9.4.7.2.1.2 a will read: "Drywell purging during plant operating modes 4 and 5 (cold shutdown and refueling) is ' unlimited and will be performed prior to personnel access as required." This.
change brings the USAR into agreement with Radiological Procedure for Drywell Entries, CPS Procedure 7200.32. This change provides more appropriate latitude for performing drywell purging.
This change clarifies why drywell purging will not be performed in Modes 1, 2, or 3, and when it will be performed in Modes 4.and 5. This change does not alter the capabilities of the Drywell Purge (VQ) System, and therefore, it does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. To use the VQ system, a makeup air supply flow puth must exist between the atmosphere, containment, and the e drywell before an exhaust air flow path is established for exhaust to the outside atmosphere. This prevents lack of make up air flow to the filtration exhaust fans which could cause excessive drywell negative pressure. Exceasive negative pressure in the drywell impacts the level of the suppression pool which may increase the possibility of the pool liquia to flow over the weir wall.          j '
, . Technical Specification Bases B3.6.5.3 states that these valves are required to remain closed where the possibility of a Loss-of-Coolant Accident (LOCA) is greatest to ensure containment /drywell boundary isolation. The possibility of a LOCA during Modes 4 and 5 are minimal; this change allows valves 1VQ01 A and IVQOlB to be open to establish an air flow path.
Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR: The VQ system will continue to perform as before and does not change
   . the VQ safety related features, therefore, the change does not create the possibility of an accident
(
I l
l 1
E

:.                                                                                         Attachment 2 to U-603276 Page 75 of 252 or a malfunction of equipment important to safety of a different type. This change incorporates
: Drywell Technical Specification Basis B3.6.5.3 and does not change the operation of the VQ
            ,  -. system; therefore, this activity does not reduce a margin of safety as defined in the basis for any aTechnical Specification; iMODULAR SUPPORT BUILDING INSTALLATION -
Document Evaluated: Coordination Plan No. CPS-97-012                             Log Number: 97-245 j
                . This change supports the temporary installation of four 24' x 68' modular office buildings on a         j permanent slab located immediately east of the Plant Support Building. These four structures will be referred to collectively as the Modular Support Buildings (MSBs). Two years is the estimated duration of this installation.~ Safety Evaluation Log number 97-230 evaluates the supporting slab.
The MSBs interface with potable water, waste water treatment, aad telephone communication isystems.' The connections to these system may result in increased volume but are well within syst'em capacities. The 12 kV " Construction Loop" provides the MSBs with their electrical i service. The Emergency Off-site Facility and the MeteorologicalTower also receive their power from the 12kV loop and is a reliable source. There is no direct interface with safety related systems, therefore, there is no increase in the probability or consequences of a malfunction of equipment important to safety. Section 3.5.2 of the Updated Safety Analysis Report (USAR) discusses External Missile Protection and Section 3.8 ~ discusses the Design of Seismic Category 1 Structures.'The closest safety related structure, the fuel building, is estimated to be about 300 feet away.- Each half of the modular building weighs in excess of 25,000 pounds and is fastened
                                                         ~
to the supporting slab. This is not a' credible missile threat to plant structures. The MSBs are not designed to withstand seismic loads; however, if the MSBs completely fail, they have no adverse
              , . impact on plant equipment. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The installation of the MSB affects the interfacing systems and the potential for missile damage, however, these are both discussed above. Therefore, there are no credible accident scenarios or malftmetions of equipment important to safety of a di'ferent type. The tornado generated missiles could compromise secondary containment. However, USAR Section 3.8.4.3 states that " secondary
                 - containment is not designed to withstard the effects of missiles." Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.                     ,
CHANGE TO THE SNUBBER FUNCTIONAL TESTING INTERVAL p                   Document Evaluated: ORM Change 20-6.                                            Log Number: 98-002 L                                                                                                                          ,
This activity changes the snubber functional testing interval specified in Section 2.4.1 of the Operational Requirements Manual (ORM) from once per 18 months to once per refueling outage            ;
in accordance with the guidance given by the first addenda to ASME/ ANSI OM-1987, Part 4.
The snubbers are designed and assumed to be an accident mitigator or to support operability of interfacing accident-mitigating systems. The failure or loss of a snubber is not an initiator for any i

O Attachment 2 to U-603276 Page 76 of 252 accident previously evaluated in the Updated Safety Analysis Repon (USAR). The accidents that
      , could be affected by this change are those involving a loss of the structural integrity of components employing snubbersf Since the snubbers still meet the same design, material, and construction standards, this activity doe's not increase the likelihood of an accident occurring.
      "Also, this change does not affect overall system performance causing the snubbers to be operated outside of their design or testing limits, causing system vibration, fatigue, or cycling outside of   ,
      - design limits, or cause a change to any system interface in a way that would increase the
  ,   _ likelihood of an accident. Some snubbers are directly connected to systems that contain radioactive material. However, malfunctions or failures of the snubbers do not necessarily involve a release of radioactive material. The proposed testing interval has been evaluated as acceptable based on component performance and industry failure data. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the.USAR. This activity does not alter the design, material, or construction standards applicable to the snubbers.
The change does not invoke any new testing or operational modes that could cause snubbers to be operated outside of their design or testing limits. Since this change is only a revision to the required interval for performing snubber functional testing, no physical equipment has been modified and no' equipment qualifications have been altered. The qualifications, design functions, _
and performance criteria of the snubbers remain the same. Therefore, this change does not increase the probability or consequences of a malfunction of equipment imponant to safety evaluated previously in the USAR. This activity revises the definition of the required interval for
                                           ~
performing snubber functional testing; there are no design or operational limits challenged or exceeded as a result of this change.1 No equipment is operat'ed differently and there are no new system interactions as a result of this change. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than
      , previously evaluated. This change is consistent with the ASME Code requirements, and is not expected to reduce snubber availability. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
DELETION OF TESTING REQUIREMENTS FOR MSIVLCS LEVEL INSTRUMENTATION Document Evaluatedi ORM Change 18-1 and CR l-97-10-107                          Log Number: 98-003
                 ~
This activity removes reference to level instrumentation as surveillance testing for the Main Steam Isolation Valves Leakage Control System (MSIVLCS) in the Operational Requirements Manual (ORM) because the MSIVLCS does not include any level instruments. The MSIVLCS controls
      ? and' minimizes' the release of fission products, which could leak through the closed main steam
      . isolation valves (MSIVs) after a Loss-of-Coolant Accident (LOCA). The system instrumentation consists of pressure, temperature, and flow instruments, which provide monitoring and control               l functions. Deletion of the ORM testing requirements for a " Channel Functional Test" and a                  ]
        " Channel Calibration". on~ non-existent level instruments does not invalidate the design criteria to which the MSIVLCS conforms. This activity does not impose additional testing requirements'on                1 the MSIVLCS, nor increase the frequency of operation of the MSIVLCS. The system does not                    ]
use level as a process variable and does not have any protective features derived from level. Since         ;
the level instrumentation does not exist, the system performance or reliability is not degraded.
i t.
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          -3..
Attachment 2 m .
1y                                                                                                 to U-603276 t                                                                                                 Page 77 of 252 i                L This activity does not make changes to system hardware and does not affect the seismic,
                 . environmental, or separation criteria. > The MSIVLCS consists of piping, valves, drain lines, blowers,' and heaters. It does not contain any tanks, or volumes for fluid collection, thus fluid level is not a process variable. The lack of a level instrument does not preclude the system or any isystem component from performing its safety function.~ Removal of the requirement to perform
                    " Channel Functional Test" or a " Channel Calibration" on a non-existent level instrument does not  -
cause any system component to malfunction so as to impair access to vital areas or prevent            )
l mitigation of a LOCA. Nor does it impair the leakage transport path to the Standby Gas Treatment Systemc . Also, this change does not compromise the design, material, or construction
: standards to which the MSIVLCS was constructed and is being maintained. Therefore, this change does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). All MSIVLCS
                 . design basis requirements and functions are maintained following this change. Therefore, this
                  . activity does not increase the consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the USAR. Removal of the ORM testing requirements for a
                  ' non-existent level instrument will not cause any credible malfunction in any piping, valves, blowers, heaters,' monitoring instrumentation, control instrumentation, interlocks, timers or any system components important to safety. Therefore, this change does not create the possibility of
                  - an accident'or malfunction of equipment imponant to safety of a different type than previously evaluated in the USAR. Technical Specification 3.6.1.8.3 discusses operability requirements for the MSIVLCS and its subsystems and components This change does not impair MSIVLCS                    j
                   . system or subsystem operability. Therefore, this activity does not reduce a margin of safety as      1
: defined in the basis for any Technical Specification.
US'AR CHANGE FOR CR NO. 1-97-09-188 LEAK DETECTION SYSTEM
                  ' Document Evaluated: USAR Change'8-129                           Log Number: 98-004 The following changes 'are being made to the. Updated Safety Analysis Report (US AR) as corrective action for Condition Repon (CR) 1-97-09-188: provide the correct reference for
                  ' sensitivity and response time of the leak detection system in Section 5.2.5.10 of the USAR, correct the reference in USAR Section 7.7.1.24.6.3.1 from Section 7.7.1.24.9 to 7.7.1.24.10, and change the unidentified leakage increase in Section 5.2.5.1.1 from "a total of 5 gpm" to "3.6 gpm or 2 gpm increase in any 24 hour period for the drywell floor drain sump flow rate and 2 gpm for the drywell cooler condensate flow rate" The first two changes are editorial in nature and do not affect any equipment important to safety,   j
                  ~therefore, these changes do not increase or create the probability or consequences of an accident      !
or malfunction of equipment important to safety. The third change corrects a technical enor by       l adding the correct terminology that was approved per. Safety Evaluation Log numbers 95-079,          I 95-080, and %-101 Revision I associated with modifications LD-027 and LD-028. The USAR               l
                   ~does not identify any. accidents, radiological parameters, or automatic actions associated with the   l alarm setpoints for unidentified leakage. Therefore, this change does not increase the probability   l or consequences of r.n accident previously evaluated in the USAR. - Leakage detection monitoring 6             i 1

g Attachment 2
* to U-603276 Page 78 of 252 of the Drywell sumps are not required for safe shutdown of the plant.' This technical change does
  - not involve any alteration to plant structures or components. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This revision specifies the correct alarm activations for both the drywell floor drain sump flow rate and the drywell cooler condensate flow rate. The specific 1  setpoints for the Main Control Room alarms are contained in modifications LD-027 and LD-028 nnd the setpoint values are n'ot affected by this change. . Therefore, this change does not create the possibility of an accident or equipment malfunction of a different type. The Leak Detection system provides an early indication'of Reactor Coolant Pressure Boundary (RCPB) leakage.
Leakage in excess of Technical Specifications allowed limits may continue to degrade the RCPB.
The acceptance limits for the Technical Specification for early identification ofleakage from the drywell floor drain sump flow rate instrumentation remain unaffected by this change. Therefore,
    . this activity does not reduce a margin of safety.as defined in the basis for any Technical
  - Specification.-
CHANGES TO CHAPTER 12 OF THE USAR L Document Evaluated: USAR Change 8-087                                             Log Number: 98-005 This change implements the following changes to Chapter 12 of the Updated Safety Analysis Report (USAR): revises a vague reference to ' radioactive pieces of steam cycle equipment' to list actual components contributing to skyshine; reflects the actual limitation on the use oflabyrinths; revises several references from ' hatches' to ' shield plugs' to more accurately reflect plant configuration; changes the descriptio~n of tanks to reflect the two types of tanks in use at Clinton      ]
{
Power Station; revises the use of area on 800'. Turbine to reflect the contamination control               I measures and use of the fuel cask'washdown area; reflects the interlock associated with Area              I Radiation Monitors on both fuel bridges; discusses the shielding limitations that apply during fuel        I
   - movement; removes the same controls on exempt sources as those applied to non-exempt sources and adds ref ence to the Operational Requirements Manual; and reflects the change in membership and chairperson of the As-Low-As-Reasonably-Achievable Committee. These changes increase specificity and provide clarity to the USAR descriptions. These changes also correct minor inconsistencies in facility description and usage of facility areas. These changes do not have a direct affect on any systems, structures, or component designs; nor do they affect operator response or equipment operation. Therefore, these changes do not increase the probability or consequences nor create the possibility of a malfunction of equipment important to safety. Since these changes do not affect operator response or any plant system, they do not change the analysis for a tank break, a dropped fuel bundle, source leak testing, or other accident analysis. Therefore, these changes ~do not increase the probability or consequences of an accident, nor do they create the possibility of a new type of accident. The Technical Specifications do not discuss the changes described in this evaluation. Technical Specification administrative z requirements for these activities continue to be met and fully implemented. Therefore, this
     . activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
p                                                                                                          j

Attachment 2 to U-603276 Page 79 of 252 RESERVE AUXILIARY TRANSFORMER PRIMARY TAP CHANGE DURING SHUTDOWN PRIOR TO RESTART
       ' Document Evaluated: Modification ~ AP-046; USAR Change 8-112                  Log Number: 98-006 Engineering Change Notice (ECN) 30690 repositions the primary winding tap of the Reserve              -
      - Auxiliary Transformer (RAT) from position #3 to position #1. This results in an increase in the tran'sformer turns ratio from 345,000/4264 volts to 362,250/4264 volts, which in turn decreases the transformer secondary voltage for a given primary voltage value. This change also alters the
      - power source for non-Class IE loads from the RAT to the Unit Auxiliary Transformers (UATs) utilizing a power backfeed through the Main Power Transformers (MPT). This allows a power feed configuration similar to that during power operaticns. This ECN is to be in place only during the plant shutdown condition until static VAR compensators are installed under modification AP-037 and other modifications are completed to permit Cycle 7 operation.
Section 8.2.2 of the Updated Safety Analysis Report (USAR) discusses distribution system undervoltage. Calculation 19-AQ-13 establishes 343 kV as the minimum acceptable grid voltage when the RAT primary tap is on position 1. Overall, this tap change improves the range of acceptable grid voltages that ensure adequate voltage availability to site Class 1E electrical buses.
The' calculation also shows that the Class lE loads have sufficient voltages for equipment to operate within their equipment ratings and fault currents are less than interrupting device capacity.
For the power backfeed, the UATs are designed to carry the auxiliary loads during power operation. Therefore, these changes do not increase the probability or consequences of an equipment malfunction previously evaluated in the USAR. Generator load rejection, evaluated in USAR Section 15.2.2, is caused by electrical grid disturbances which result in significant loss of electrical load on the generator, failure of any MPT, UAT fault detect:on, overcurrent, or differential overcurrent relay trip. During plant shutdown, there is no generator load, however, the MPTs are the source for backfeeding the UATs from the 345 kV system. The MPTs are designed to carrj the full generator output; and therefore, have sufficient capacity to supply the UAT loads during the backfeed operation. USAR Section 15.2.6 evaluates the loss of alternating current (AC) power. The RAT is designed to carry the startup load in addition to the total coincident Loss-of-Coolant Accident'(LOCA) load. Under this ECN, the RAT carries less than              j 4
its design capacity since the non-Class IE loads are fed from the UATs. Therefore, this change -        ;
      - does not increase the probability or consequences of an accident previously evaluated in the USAR. Voltages are within equipment ratings and fault currents remain below interrupting ratings of Class IE protective devices. Each UAT is designed to carry one-half of the full load auxiliary requirements. Therefore, this change does not create the possibility of an accident or
       . equipment malfunction of a different type. Technical Specification Bases B3.3.8.1 defines the bases for the degraded voltage relays; these changes are within those limits. Technical Specification Bases B3.8.1 define the bases for the offsite and onsite power circuits; these changes help ensure these assumptions are met: Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.

l Attachment 2            ,
to U-603276
                    ,                                                                                Pa3;c 80 of 252
  ,                  RHR B HEAT EXCHANGER TESTING
                   ' Document Evaluated: CPS No; 2602.01D0032 R/l                                    Log Number: 98-007 l
J~
                  ' This change addresses the use of the Emergency Core Cooling System (ECCS) pumps in the test mode to heat the suppression pool in preparation for heat exchanger tests. The pumps provide              -
heat to warm the pool and the higher temperature in the pool provides a larger temperature across             i the heat. exchanger and more accurate test data and results. The suppression pool temperature and ECCS system operating parameters will remain within established limits. This change revises Updated Safety Analysis Report (USAR) Section 6.3.2.6 to include using the ECCS pumps for extended periods of time to heat the suppression pool.
L The equipment is designed for a 40 year life; the additional use of 240 hours has been evaluated for impact on equipment life and found to be acceptable. . The pump and motor performance are monitored monthly and any degradation due to wear would be found. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. Extra run times could add wear to the pump over a long period of time but does not cause pump or pipe failure problems or other accidents.
Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The pumps and motors have been qualified for the life of the unit and                 I' their performance is monitored as required by Technical Specifications and the Maintenance Rule.
Therefore, this change does not create an accident or equipment malfunction of a different type.
The life extension evaluation concludes that the systems are still able to perform their design functions; therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
4 ECCS SUCTION STRAINER ~
Activity Evaluated:L Modification M-083,                                      Log Number: 98-008R/l TS Bases Change No. BL-98-014, and USAR Change 8-146 Modification M-083 removes the individual suction strainers connected to each Emergency Core
                  - Cooling System (ECCS) and Reactor Core Isolation Cooling (RCIC) system penetration piping and installs a large passive strainer that rests on the floor of the suppression pool. The new strainer is connected to the penetration where the previous strainers were removed. The purpose of this new strainer is to alleviate the concern that the current suction strainers are marginally sized for the postulated Loss-of-Coolant Accident (LOCA) generated debris loading. The new
                   . strainer has been designed to. withstand postulated seismic, hydrodynamic and other applicable loads, rad to minimize clogging under postulated conditions. Similar code and construction requirements were used to build the new strainer as were used with the original strainers. The new strainer has been designed to exceed the functional requirements of the origina! strainers.
                                             .A

4 s                                                                                         Attachment 2 to U-603276 t-Page 81 of 252 -
The ECCS strainers _are involved in the mitigation of accidents and transients, and are not event initiators. Updated Safety Analysis Report (USAR) Secti,a 63.2.5 states,"A single failure
         . analysis shows that no single failure prevents the starting of the ECCS when required, or the delivery of coolant to the reactor vessel. No individual system of the ECCS is single failure proof with the exception of the ADS, hence it is expected that .agle failures will disable individual systems of the ECCS. The most severe effects of single failures with respect to loss of equipment L
l          occur if the LOCA occurs in combination with an ECCS pipe break coincident with a loss of                  ,
L          offsite power." Since this modification interfaces with the ECCS and RCIC suction lines only, it does not affect anything that could cause a LOCA or an ECCS pipe break coincident with a loss
                                                                                                                      )
        - of offsite power. This change does not degrade the performance of any bounday that mitigates                J
;          dose to the public. The change canr,at result in a pathway for release of radioactive material, because the design is such that the loading on the containment and containment penetrations associated with the strainer is within the original design limits for the containment. A postulated       ,
l           failure of the new strainer would affect the operability of the ECCS and/or RCIC system.
!          However, the new strainer has been designed,such that a failure would not result in loss of more L
than one functional group of the ECCS.which is consistent with the original strainer design.
Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. Malfunctions associated with the new ECCS/RCIC suction strain s, which could lead to failure of the ECCS/RCIC system or containment heat removal system function, are clogging of the strainer or structural failure of the strainer. The results of structural
: failure of the new strainer would be the eventualloss of an entire division of ECCS. Dased on analyses, the ' probability of the failure of the new large passive suction strainer is no higher than that for the original strainers. The new strainer was designed and built using similar code and construction requirements as the original strainers, and has been designed to exceed the functional requirements of the original strainers. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. ' This change does not change the overall ECCS or RCIC system performance or reliability, nor does it cause the systems to be operated outside of their design basi's .or test iimits.
Since the new strainer is passive, no operator actions are required, and thus, does not increase the possibility of operator error. Therefore, this change does not create the possibility of an accident
          'or malfunction of equipment of a different type than previously evaluated in the USAR.
This modification does not reduce the margins inherent in the acceptance criteria of 10CFR50.46 l since the design of the new strainer allows and enhances the ability of the ECCS to satisfy these requirements by maintaining the Net Positive Suction Head for the ECCS pumps within their design requirements; nor does it modify the response or reduce the effectiveness of the RCIC system. Technical Specifications Bases 3.6.2 was revised to address the reduction in the minimum volume of water available in the suppression pool (as a result of the increased size and displaced volume of the new strainer), however, a change in the Technical Specifications for l
p         ! suppression pool level limits was not required, since maximum and minimum suppression pool l            levels as required by Technical Specifications are unaffected by this modification. No new design
!'           controlling loads have been introduced in the suppression pool due to the addition of the strainer.
L            Therefore, this' activity does not reduce a margin of safety as defined in the basis for any Technical   i Specification.

Attachment 2                                '
to U-603276 Page 82 of 252
: TRAINING FOR ESCORTED RADWORKERS AND TOURIST ESCORTS
                                        , ' L Document Evaluated: USAR Change 8-095                                                              Log Number: 98-010 This change revises Updated Safety Analysis Report (USAR) Section 13.2.1.1.1 Paragraph R by '
                                                           - replacing exposure limits _with exposure levels. . Exposure levels are levels at which dose                                            -
extensions are needed and specific controls will be applied over ' additional dose. - In the past, exposure levels were defined as limits. By being limits, th'ey were approached carefully and in fact resulted in other control levels being established to avoid exceeding the exposure limits. Specific controls shall be applied as needed to ensure the individual involved does not exceed the authorized exposure level without prior authorization.fOnce the exposure level is reached, further exposure of the individual should be avoided. _ The change is a change to administration of the control of radiation dose to personnel. It does not bolve operation, maintenance, or design of -
                                                            ; equipment or its environment.' Therefore, this change.does not increase or create the probability .
or consequences of an accident or malfunction of equipment important to safety. The Technical
                                                              . Specifications do not discuss radiation exposure dose controllevels for escorted radwoikers.
Therefore, this =tivity does not reduce a margin of safety as defined in the basis for any Technical .
Specification.
CHANGE IN START TIME FOR DIESEL GENERATOR ROOM 1B SUPPLY FAN J1VD01CB
                                                                                                      ~
Document Evaluated: Temporary Modification 98-001                              Log Number: 98-011 Temporary Modification 98-001 replaces the Class lE Agastat relay (1UAYVD508A) with a new Class lE Agastat adjustable time delay relay. This Temporary Modification is for Division II only. This new relay delays the start of the Diesel Generator (DG) Room Heating, Ventilation, and Air Conditioning (VD) Supply Fan (IVD01CB) until after the Degraded Voltage Relay timer times out. This ensures that the DG Room IB Supply Fan starts if offsite voltage is at the L                                                               minimum expected and that the VD system performs its design basis function. This Temporary .
Modification is only in place during Mode 4. ; This activity does not change the safety function of the rd.y. An evaluation shows that delaying the start of fan 1VD01CB results in no adverse
: impact on the DG Room temperature. ; The rated mechanical life of both the original relay and the new relay ~is 100 million operations, and they both have the same contact failure characteristics.
The probability of the failure of the new time delay relay is the same as the original relay. The consequences of failure are the same for both relays -- a failure of the VD supply fan to operate
                                                           . when required. Therefore, this change does not increase the probability of a malfunction of fequipment important to safety previously evaluated in the Updated Safety Analysis Report
                                                            ; (USAR). This relay and its associated circuit are unrelated to any assumed accident initiator
                                                              . discussed in Chapter 15 of the USAR. USAR Table 9.4-10 " Diesel Generator Facilities
                                                            ~ Ventilation System Faihire Analysis" evaluates the failure of the DG Room IB Supply Fan to start. The loss of a ventilating fan does not affect the safe shutdown capability of the station, since the plant has 'two independent Diesel Generators each of which is capable of safely
                                                                                                        ~
mitigating an accident. Also, the analysis for the Temporary Modification shows that there is e

.                                                                                     Attachment 2 to U-603276             4 Page 83 of 252 adequate power from the offske sources and the DG should not be needed. Therefore, this tctivity does not increase the probability or consequences of an accident previously evaluated in       j the USAR. The new time delay relay could drift from its setpoint. To alleviate this concern, the         l new time delay relay was checked by verifying that the delay time between the DG start signal and        l the fan breaker closing is within the allowable tolera. ces. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type    -
than previously evaluated. This activity does not affect any Technical Specification safety limits, limiting conditions for operation, or design limitations associated with the VD system. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.                                                                                           I i
l REVISE SAR DESCRIPTION TO CONFORM WITH EXISTING MAIN STEAM SRV DESIGN CONFIGURATION Document Evaluated: USAR Change 8-089                                           Log Number: 98-012 i
This activity includes the followtng facility changes: revice Updated Safety Analysis Report (USAR) Section 7.3.1.1.1.4.2.6 to correct the number and type of Safety Relief Valves (SRVs) undergoing changes in their opening and closing setpoints and clarify that when reactor pressure reaches any of the normal relief setpoint levels, the low-low set logic is automatically sealed in; revise USAR Section 5.2.2.2.3.2 to clarify that any SRV relief mode actuation will override normal setpoints on five of the SRVs for low-low set re2-f function; revise US AR Section              q 110.3.4 to clarify that relief actuation of any SRV will arm the low-low set relief function by deleting reference to the lowest and second lowest relief setpoint group of valves arming the low-      j low set relief function.                                                                                j l
This change revises the description of the SRV low-low set relieflogic function in the USAR to          l accurately reflect the original, existing design without any hardware change. This change does          !
    . not impact the minimum number of operable SRVs and low-low / Automatic Depressurization                  l System (ADS) features of the SRVs as described in Chapter 15. This change does not impact any           l systems, structures, or components in any of the systems described in Chapter 6 as forming the          l basis to mitigate the consequences of a postulated design basis accident. The clarification of the low-low set point for the SRVs does not alter the affect of SRV blow down into the suppression pool and radiological impact on the environment inside or outside the containment. Therefore,          q this change does not increase the probability or consequences of an accident previously evaluated       j in the USAR. This change does not impact the ability of the SRVs to perform their design and            j Technical Specification intended function. There is no change to the equipment, testing procedures, or testing requirements. Therefore, this change does not increase the probability or consequences of an equipment malfunction previously evaluated in the USAR. There exists no            ^
l hardware change and no change not previously evaluated or assumed in the safety analyses.
Therefore, this change does not create the possibility of an accident or equipment malfunction of a    ;
difTerent type. This change brings the USAR in compliance with current plant configuration and         i Technical Specification requirements. Therefore, this activity does not reduce a margin of safety     !
as defined in the basis for any Technical Specification.                                               l l
l s

Attachment 2 to U-603276 Page 84 of 252 EVALUATION TEST OF DIVISION 2 NSPS INVERTER Document Evaluated: CPS No. 2813.00                                             L.og Number: 98-013 Clinton Power Station (CPS) Procedure 2813.00, " Evaluation Test of IC71-S001B Inverter," is a new procedure for performing evaluation testing of Division 2 Nuclear System Protection System (NSPS) inverter. This procedure helps identify the cause of the inverter (IC71-S001B) transferring the Division 2 NSPS bus (IC71-P001B) from the normal inverter power supply to the alternate power supply for e identifiable reason. This procedure allows stopping the test at any time and returning the in         and bus loads to a safe and stable condition. During the performance of this procedure,1C71-S001B and IC71-P001B are out of service with the appropriate Improved Technical Specification and Operational Requirements Manual actions in place. The normal DC feed to the inverters protected by two breakers, any fault current or failure of the equipment must first pass through the equipment and both breakers. Prior to returning the equipment to an OPERABLE status, all test equipment will be returned to design configuration.
Therefore, this change does not increase or create the probability or consequences of an accic'ent or a malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report. This ac ivity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
MODIFICATION TO SX SYSTEM TO ADDRESS NRC INFORMATION NOTICE 92-18 Document Evaluated: Mod SX-048                                                  Log Number: 98-014 Modification SX-048 alters the wiring for Motor Operated valves in the Shutdown Service Water (SX) system in response to Nuclear Regulatory Commission (NRC) Information Notice 92-18.
This modificaticri reconfigures the control circuits to allow the limit and torque switches of the valve to open the circuit and prevent the valve actuator from being continuously energized. The modification assures that the valves will not be damaged due to potential" hot shorts" caused by a fire in the Main Contro1 Room (MCR). The response of these valves to the operation of their respective control switches is exactly the same. All of the interlocks including both accident response and nornial operating functions have not changed. Therefore, this modification does not increase the probability or consequences of a malfunctica of equipment important to safety             1 previously evaluated in the Updated Safety Analysis Report (USAR). Single 'ailures of the SX system have been evaluated, but with the loss of one division, the other dir i would be available to perform the intended functions. The aherations do not alter oi T 3 duce any interactions with other components. Therefore, this change does not increa., he probability or consequences of an accident previously eva!uated in the USAR. The changes to the valve controi circuitry are internal to the circuitiy and do riot involve any component or equipment which is not already interlocked with the various valves. This modification does not introduce any interactions with other components; therefore, this change does not create an accident or equipment malfunction of a differen.t type. This modification protects the valves from damage as a result of a
  " hot short" in the MCR and ensures the valves can be operated remotely. This activity does not l
l I

4
* Attachment 2 to U-603276 Page 85 of 252 impact any Technical Specification safety limits limiting safety system settings or limiting conditions for operation. Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
JUSTIFICATION FOR LEAVING TEMPORARY TEST EQUIPMENT INSTALLED IN RHR B PUMP ROOM Activity Evaluted: Temporary Modification 98-010                                 Log Number: 98-015 Temporary Modification 98-010 installs instmmentation on Residual Heat Removal (RHR) Heat Exchanger IB piping for the purpose of verifying heat exchanger performance in support of Nuclear Regulatory Commission Generic Letter 89-13 requirements. The instrumentation installation was evaluated to not create auy seismic or transient combustible loading concerns via the temporary modification. This change will not have an adverse impact on the RHR "B" system. For a postulated fire in Zone A3-a, no credit is taken for equipment or equipment with electric cables routed through this zone in safe shutdown of the plant. Credit for achieving and maintaining safe shutdown is via Method 3 as described in the Safe Shutdown Analysis. Fire zone A3-a is rated as " low" The total weight of the wiring insulation has been determined to be less ihan 30 pounds and the additional fire loading is minor and does not change the rating. The thermocouples extension wires are isolated to Fire Zone A3-a. The amount of materiel added is not sufficient to warrant concern of the fire spreading to another zone. The voltage generated by the thermocouples and transmitted by the wires is very low. These wires were not routed close to a heat source. Consequently, this does not introduce an internal or external heat source. This Temporary Modification does not interact with other systems. Once in place, this measuring equipment performs a passive monitoring function. The weight contribution from the installed thermocouples is negligible and the presence of the monitoring equipment is acceptable based on seismic evaluation. Therefore, this activity does not increase the probability or consequences of    ,
an accident or malfunction of equipment important to safety previously evaluated in the Updated       l Safety Analysis Report (USAR). The hazards of the temporary materiah, are that PVC expels toxic gas during burning. Whether this small amount of material will burn more readily or not will not cause the fire to spread to an additional fire zone or cause more damage than postulated for this zone. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This activity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting i
conditions for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
MODIFICATION TO RH SYSTEM TO ADDRESS NRC INFORMATION NOTICE 92-18 Document Evaluated: Modification RH-047                                        Log Number: 98-016 Modification RH-047 alters the wiring for Motor Operated valves in the Residual Heat Removal (RH) system in response to Nuclear Regulatory Commission (NRC) Information Notice 92-18.
This modification reconfigures the control circuits to allow the limit and torque switches of the

Attachment 2           ~
to U-603276 Page 86 of 252 valve to open the circuit and prevent the valve actuator from being continuously energized. The modification assures that the valves will not be damaged due to potential" hot shorts" caused by a fire in the Main Control Room (MCR). The response of these valves to the operation of their respective control switches is exactly the same. All of the interlocks including both accident response and normal operating functions have not changed. Therefore, this modification does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). The USAR evaluates the malfunctioning of the RH system, but does not identify specific component failure. The changes do not increase the possibility of a failure of the individual valves, therefore, this change does not increase the probability of an accident previously evaluated m the USAR. The changes to the valve control circuitry are interint to the circuitry and do not involve any component or equipment which is not already interlocked with the various valves. This modification does not introduce any interactions with other components, therefore, this change does not create an accident or equipment mdfunc: ion of a different type. This modification protects the valves from
. damage as a result of a " hot short" in the MCR and ensures the valves can be operated remotely.
The Technical Specifications describe the need for either two trains of RH or one train of RH plus another injection system. This change does not invalidate this basis; therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
IMPLEMENT CORRECTIVE ACTIONS AND IMPROVEMENTS TO THE SAFETY EVALUATION PROCESS AT CPS Document Evaluated: CPS No. 1005.06, " Conduct                                    Log Number: 98-017       ;
of Safety Reviews" Clinton Power Station (CPS) Procedure 1005.06, " Conduct of Safety Reviews," is the site procedure for conducting evaluations for changes, tests, and experiments to ensure compliance with Title 10, Part 50, Section 59 of the Code of Federal Regulations and other regulations. This change revises CPS No. 1005.06 to incorporate corrective actions for several condition reports, to implement a previously approved advance change, and other changes resulting from reviews of the Safety Evaluation Program. The functions and responsibilities for the Facility Review Group, Nuclear Review and Audit Group, and the Independent Safety Engineering Group as described in the Updated Safety Analysis Report (USAR) are altered by this change. The changes are programmatic in nature, and do not directly or indirectly affect failures of plant equipment.              I Therefore, the changes do not increase the probability or consequences of a malfunction of                 i equipment important to safety previously evaluated in the USAR. The changes affect how and when to perform a Safety Evaluation, they do not impact accident initiators or assumptions included in the accident analysis. Therefore, these changes do not increase the probability or consequences of an accident previously evaluated in the USAR. Since no new failure modes or mechanisms for any component or system are introduced, this change does not create the possibility of a different type of accident or equipment malfunction. As stated above, these changes are programmatic in natcre and do not impact any barriers or any other acceptance limits.
Therefore, these changes do not reduce a margin of safety as defined in the basis for any Technical Specification.

.                                                                                  Attachment 2 to U403276 Page 87 of 252 REROUTING OF DIVISION 2 CABLES FROM FIRE ZONE CD-lO Activity Evaluated: ECN 30622 and USAR Change 8-094                            Log Number: 98-018 Engineering Change Notice (ECN) 30622 reroutes Division 2 cables 1NB65E,1RH79E and 1RP76E to a location not requiring use of Thermo-lag insulation. Currently, the cab'es are routed through Fire Zones CB-6A, CB-2, CB-lG, and CB-6B. Thermo-lag insulation is used in Fire Zone CB-lG to protect the Division 2 cables from a fire in this zone. The new routing bypasses Fire Zone CB-lG and eliminates the need for Thermo-lag insulation for the cables. The Updated Safety Analysis Report (USAR) is being revised to eliminate all references to the use of Thermo-lag insulation in Fire Zone CD-lG and to remove all references to routing of Division 2 cables in the zone. The instruments associated with the cables are required to function in support of various accident scenarios. Post-modification testing assures circuit integrity has been retained, thus the instruments function as before. The modification was performed in accordance with approved procedures and specifications for safety-related equipment. Thermo-lag has been determined to not be fully effective in preventing fire damage, and therefore, can not be relied upon to preserve the safe shutdown in Fire Zone CB-lG. Eliminating the need for Thermo-lag protection of safety related cables enhances the plant design and helps ensure that the cables and associated instruments function properly in the event of a fire. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. All equipment imponant to safety continues to function as designed in support of plant operation and safe shutdown. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The rerouting of the Division 2 cables has no affect on the function of the associated instruments, instrument loop integrity, or instrument setpoints.
Appropriate post-modification testing ensures that the instruments continue to ftmetion as designed, in order to maintain margins of safety. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
DIESEL GENERATOR (DG) ROOM MAXIMUM & MINIMUM TEMPERATURES Activity Evaluated: ECN 30716 and USAR Change 8-288                            Log Number: 98-021 Engineering Change Notice (ECN) 30716 revises the results of extreme outdoor conditions in various design and licensing documents. With the Diesel Generator (DG) running concurrent with outside air temperature of 112 F, the DG Room temperature will approach 140 F for no more than 12 hours. When outside air condition drops to as low as -22 F, the DG Room can be maintained at a temperature greater than 65 F by the heat generated by the DG.
The limiting condition for this activity is the acceptance of 140 F temperature inside the DG rooms due to extremely high outside temperature of 112 F. This condition could initiate failure of components inside the DG rooms and shutdown the DG. However, it was determined th st the          i short term excursion of temperatures would not affect the operability of the DG and associated      l components and would not prevent the DG and its ventilation system (VD) from performing their       l i
i

Attachment 2 ~       .
to U-603276
(                                                                                                 Page 88 of 252 f ' safety related functions. Also, high outside air temperature could initiate fire in the DG Oil Tank rooms.' The flash point of type 2D oil in the Day and Storage tanks is 125 F. However,
: temperature inside the Oil Sturage Tank rooms and Day Tank room will be 117 F and 113 F, respectively. The heat added by the fan to the exhausted air will not reach the oil flash point. The DG and associated system continue to perform their safety functions and will support actions described or assumed in the accis snalysis for mitigating the effect of the accident or transieri              -
Therefore, this activity does not increace_the probability or consequences of an accident or e malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). No failure can be postulated by implementing this activity other than those
        . failures already identified. Therefore, this change does not create the possibility of an accident or
         - malfunction of equil ) ment important to safety of a different type than previously evaluated. This activity does not change the design requirements of the DG or VD systems, nor the way they operate.' The DG loadings and fuel consumption remain well below the original values which were bounded by previous engineering evaluations. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
DIESEL GENERATOR ROOM SUPPLY FAN AUTO-START T!JdE DELAY Document Evaluated: Temporary Modification 98-007                                       Log Number: 98-022 Temporary Modification 98-007 replaces the Class 1E Agastat relay (1UAYVD507A) with a new Class IE Agastat adjustable time delay relay. This Temporary Modification is for Division I only.
This new relay delays the start of the Diesel Generator (DG) Room Heating, Ventilation, and Air Conditioning (VD) Supply Fan (1VD01CA) until after the Degraded Voltage Relay tirner times out. This ensures that the DG Room 1 A Supply Fan starts ifoffsite voltage is at or above the minimum expected and that the VD system performs its design basis function. This Temporary -
        . Modification is only in place during Mode 4. The DG Room 1 A Supply Fan and VD systems are accident mitigation support systems. They are not assumed initiators of any accidents described in the Updated Safety Analysis Report (USAR). Therefore, this activity does not increase the
        . probability or consequences of an' accident previously evaluated in the USAR. This activity does not change the safety function of the relay. An evaluation shows that delaying the start of fan
         ' IVD01CA there is no adverse impact on the DG Room temperature. The rated mechanical life of both the original relay and the new relay is 100 million operations, and they both have the same
        ' contact failure characteristics. The probability of the failure of the new time delay relay is the same as the original relay. The consequences of failure are the nme for both relays -- a failure of the VD supply fan to operate when required.' Therefore, this change does not increase the probability of a malfunction of equipment important to safety previously evaluated in the USAR.
The new time delay relay could drift from its setpoint. The failure of the time delay function could cause the VD fan to stan immediately on the Loss-of-Coolant Accident initiation signal,
         -possibly causing the voltage at the 4kV bus to remain too low for the degraded voltage relay to reset, resulting in the 4kV bus transferring to the onsite source. The required safety related o
equipment would then stan from the DG and continue to function as designed. However, these failure modes are present in the existing relay. Therefore, this change does not create the-possibility of an accident or malfunction of equipment important to safety of a different type than

4
     .                                                                                        Attachment 2 to U-603276 E     '''                                                                                  Page 89 of 252 previously evaluated. This activity does not affect any Technical Specification safety limits,
             ' limiting conditions for operation, Lor design limitations associated with the VD system: Therefore,
             ' this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.               -
             ! MODIFICATION TO RI SYSTEM TO ADDRESS NRC INFORMATION NOTICE 92-18                                     -
              - Document Evaluated:: Modification RI-049                                      Log Number: 98-023
                                                                  ~
Modification RI-049 alters the wiring for Motor Opertted valves in the Reactor Core Isolation Cooling (RI) system in response to Nuclear Regulatory Commission (NRC) Information Notice 92-18. This modification reconfigures the control circuits to allow the limit and torque switches of the valve to open the circuit and prevent the valve actuator from being continuously energized. The modification assures that the valves will not be damaged due to potential" hot shorts" caused by a fire in the Main Control Room (MCR).- The response of these valves to the operation.of their respective control switches is exactly the same. All of the interlocks including both accident response and normal operating functions have not changed. Therefore, this
             ._ modification does not increase the probabilr , or consequences of a malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). The RI system is not required to meet the single-failure criterion, however, the RI system meets the single-failure criterion on a network basis in conjunction with High Pressure Core Spray. The changes do not alter or introduce any interactions with other components. Therefore, this change
              .'does not increase the probability or consequences of an accident previously evaluated in the USARE The changes to the valve control circuitry are internal to the circuitry and do not involve any component or equipment which is not already interlocked with the various valves. This Emodification does not introduce any interactions with other components, therefore, this change            i does not create an accident er equipment malfunction of a different type. This modification protects the valves from damage as a result of a " hot short" in the MCR and ensures the valves can be operated remotely. This activity does not impact any Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation. Therefore, this activity does
              - not reduce a margin of safety as defined in the basis for c.ny Technical Specification.
              ' CLARIFICATION TO RG 1.106 REGARDING MOV THERMAL OVERLOAD PROTECTION Document Evaluated: USAR Change 8-096 and -                                   Log Number: 98-024         i ORM Change 18-2 This change revises the Updated Safety At9ysis Report (USAR) to state that Motor Operated Valve (MOV) thermal overload protection circuitry may be temporarily placed into service for
               . valve repositioning during normal operation, maintenance, and testing. It also provides guidance         ,
in the Operational Requirements Manual (ORM) regarding the length of time the bypass circuitry           l should be allowed to remain out of service. Compliance with Regulatory Guide (RG) 1.106 is               l provided by c'ompliance with the requirements of ORM Section 3.5.2.

p?                                                                                                                                         )
Attachment 2                 ,
to IM03276 Page 90 of 252 During valve repositioning, the thermal overload devices protect MOV motor windings against excessive heating. By placing thermal overload protection in service during testing or
                         - maintenance for shon time durations, the potential for MOV damage or failure during normal operation is reduced. When the valve is required to perform its safety function, the thermal overload protection will be bypassed.' Therefore, this change does not increase the probability or consequences of a malfunction of_ equipment imponant to safety previously evaluated in the                    -
                         ;USAR.' It will not increase the probability of a loss of shutdown cooling event as discussed in USAR Chapter 15. Only one of the two shutdown cooling suction valves have been provided with a thermal overlo'ad bypass in the open direction. If the valve tripped during opening when going into normal shutdown, it could be reset, bypassed and the valve then opened. Also, manual action could be taken to open the valve. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. This change does not introduce any new failure modes. It helps protect the equipment from potential malfunctions that may occur during these times. - Therefore, this change does not create an accident or equipment malfunction of a different type. The Technical Specifications do not have any specific
                         - requirements rega ding MOV thermal overload protection. Thermal overload protection will be bypassed during ac ident conditions, therefore, compliance with Regulatory Guide 1.1% is maintained. Therefore, this activity does not reduce a margin of safety as dermed in the basis for
                         = any Technical Specification.
                                                                    ^
CLOSED COOLING WATER (CCW) CORROSION CONTROL PROGRAM Document Evaluated: CPS 1830.00 Rev. 2, CPS 6001.01 Rev. 25, Log Number: 98-025 CPS 3223.01/D001 Rev. 8 These procedure revisions allow addition ofchemicals . > the Closed Cooling Water (CCW) system that were not included in previous revisions of these procedures. Some of these chemicals are already added as components blended into the nitrite corrosion inhibitor. The Updated Safety
                         . Analysis Report (USAR) refers to corrosion inhibitors for several CCW systems, but does not elaborate on the chemical constituents necessary for an effective inhibitor. These procedure revisi6ns also allow the use of biocides, which inhibit microbe induced corrosion and other
                          ~ deleterious effects of bacteria? The addition of biocides to these systems constitutes a change to a.
                         - procedure as described in the USAR. The biocides recommended by EPRI and by BetzDearborn are Gluteraldehyde and Isothiazolone. Both of these products are categorized as minimal reactivity hazards, have a flash point greater than 200 F, and ao corrosive in the supplied concentration; Both products have a low volatility of s 18.0 mmHG.' If there were a spill, the concentration in the air would be minimized by the low volatility. The low air concentration would minimize the impact on control room habitability. A quantitative evaluation of the potentialimpact on control room habitability is not required, because the quantity of this material at Clinton Power Station is limited to 100 pounds within 0.3 miles of the control room.                        ;
                         . Therefore, it iaeets the requirements of Regulatory Guide 1.78. There are two users of these biocides, radwaste and chemistry, e.nd they are limited to five gallon quantities each. Both Gluteraldehyde and Isothiazolone are non-corrosive to the CCW metals at the recommended use levels. Procedures specify chloride limits for each CCW system and do not allow the addition of 9
is.  . . . . _ ..-, . ,              ,

m       ,
        ..                                                                                            Attachment 2 s
to U-603276 Page 91 of 252 e
: l.                                                                                        .
l                 i the chemical unless'the calculated chloride concentration is less than the limit. If a malfunction -      )
were to occur, the water in the system could be released to the liquid radwaste processing system.        I
                   ' Prior to any discharge to the' environment, the CCW waste water will be analyzed to verify that          l f the biocide ha.s decompose'd such that the biocide is non-detectable in the waste water. Proper i usage of the chemical should not result in any deleterious effect to any CCW system component
: l.                                                                                                                 .
l                  : and should contribute to overall reliability of the CCW system: Therefore, this change does not '      -
i                   increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The Drywell Cooling System and Plant Chilled Water
                                               ~
l                    System fluid enters the drywell, so potentially the additives could be activated by neutrons in the
    <             . drywell. The new chemical elements that would be added to the systems by using Gluteraldehyde L                    and Isothiazolone are sulfur, bromine, and magnesium. Review of The Chart of the Nuclides identifies radioactive neutron capture daughter products from these elements. The daughter with
!                  ; the longest halflife is sulfur-35, however, it does not emit any gamma rays in its decay. Next is .      {
                  - biomine-82 with a 1.47 ' day half-life. Because the neutron flux in the drywell is low, daughters '       i with half-lives of 1.47 days or less will not buildup to' pose a significant radiological concern.
l                    Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This activity does not affect any Technical Specification safety limits, limiting' conditions for operation, or design limitations.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical
                   . Specification.
l l                    SIMULTANEOUS VERIFICATION OF THROTTLE VALVES Activity Evaluated:- US AR Change 8-100                                           Log Number: 98-027 This activity revises the Updated Safety Analysis Report to reflect the appropriate method of second verification of throttled valves.: This change does not alter the physical configuration of      ;
the plant, nor does it alter the way these valves are operated. The administrative activity being controlled by this change affects all plant systems, both safety and non-safety related. However, this change only reflects the most positive method of second checking, and does not degrade the performance of any equipment.: This change does not introduce a new/ credible fr 2ure mode, because the position'of the throttle valve is unchanged by this administrative action. System l
performance is enhanced by making the method for second verification of throttle valves throttled to a pre'-determined ' position clear. This method ensures the valve is not mispositioned by the second checker. Therefore, this change does not increase or create the probability or consequences of an accident or malfunction of equipment important to safety. The Technical Specifications do not' address the n>ethods used to verify the positioning of throttled valves. Since
                   . this activity does not introduce any new valve alignment or system configuration, it does not
                    . reduce a margin of safety as defined in the basis for any Technical Specification.
L                                            '
L

1 Attachment 2             ,
to U403276                   l Page 92 of 252             !
TEST OF DIVISION 2 INVERTER Document Evaluated: CPS Procedure 2813.01                                    Log Number: 98-028 (duplicate log number)
Clinton Power Station (CPS) Procedure 2813.00, " Evaluation Test of IC71-S001B Inverter," is a special test procedure for performing evaluation testing of Division 2 Nuclear System Protection System (NSPS) inverter. The test equipment being used to evaluate the inverter is not shown on
* E02-lRP99-102, which is incorporated by reference in Updated Safety Analysis Report (USAR)
Table 1.7-1. While the test eqCpment is being installed, the inverter is " Tagged Out", and CPS Procedure 3509.01C002, " Division 2 NSPS Bus Outage Checklist," is implemented to prevent the loss of the Division 2 NSPS bus from causing loss of shutdown cooling. While the test equipment is installed, the inverter, as well as the rest of the plant equipment, will operate normally.
The inverter is not relied upon to perform any safety function. The addition of the test equipment does not disable the protective features of the breakers. This procedure does not prevent the NSPS bus from responding as it normally would. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. Loss of the inverter causes the power to transfer from the DC source to the alternate AC source. The test does act prevent this transfer. Also, CPS 3509.01C002 puts barriers in place that prevent loss of shutdown cooling that could occur if the Division 2 NSPS bus were to be inadvertently lost. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The test equipment is passive, consisting ofclamp on current detectors and high impedance voltage probes. The test equipment and test points have been chosen to prevent any impact on the inverter or the connected safety equipment. Therefore, this change does not create the possibility of an accident or equipment               ,
malfunction of a different type. Technical Specification 3.8.8 requires only one inverter in MODE         j
: 4. During the testing, the Division 2 NSPS inverter is considered INOPERABLE; however, the                l Division 1 NSPS inverter will continue to operate in a normal manner. Therefore, this activity             {
does not reduce a margin of safety as de. fined in the basis for any Technical Specification.             j CPS ORGANIZATIONAL CHANGE                                                                                  '
Document Evaluated: USAR Change 8-103                                          Log Number: 98-028 and ORM Change 20-2                                  (duplicate log number) l This change revises Chapter 13 of the Updated Safety Analysis Report (USAR) and Section 6 of the Operational Requirements Manual to incorporate a new Clinton Power Station (CPS) management and operating organization. In addition, administrative changes that revise titles and references in other USAR sections, in support of the organizational changes, are also included.
This change does not change any structures, systems, or co'mponents. CPS remains committed to American National Standards Institute's (ANSI) standard ANS 3.1-1978, "American National                    l Standard for Selection and Training of Nuclear Power Plant Personnel." This organizational                 l

.-                                                                                         Attachment 2 to U-603276 Page 93 of 252
        ' change does not affect the responsibilities of the operating crews. The organizational changes do      -
not impact the ability of the operating crew or maintenance organizations to mitigate an accident.
        ' Therefore, this change does not increase or create the probability or consequences of an accident or equipment malfunction. Tebhnical Specifications 5.2 and 5.3 address manning and_                        ,
qualifications. CPS is 'still in compliance with ANSI standard ANS 3.1-1978, and therefore, still meets the Technical Specification requhements. Therefore, the organizational changes do not              -
reduce a margin of safety as defined in the basis for any Technical Specification.
RESOLUTION OF CRs'l-98-01-452 AND l-98-02-060 Document Evaluated:- _ECN 30727 and USAR Change 8-108                            Log Number: 98-031 Engineering Change Notice (ECN) 30727 rewires the Direct Current (DC) feed to Division 1 Diesel Generator (DG) Panel IPL12JA to reduce the voltage drop between panels IDCl3E and
                      ~
IPL12JA and to reduce the current carried by switch SB-9. The change uses two existing cables and eliminates another cable. The design change separates the existing DC feed into two feeds,'
        - one for control power and one for motors. The shoner circuit lengths established by this design L reduce the overall impedance and line resistance losses and improve the voltages seen at the DG panel. The equipment energized by these feeds is limited to Division 1 equipment associated with starting and operating the Division 1 DG. By separating the control circuit feed for the panel from the DC motor feed, the design eliminates the dependence of the control circuit voltage on the starting / running status of the motors and assures consistently higher voltage for the control devices. This activity does not change the DC bus that feeds the panel, the only change to the equipment being fed is an additional alarm relay. The additionalload represented by this relay is too small to impact the battery capacity. The new cable and circuit breaker added by this design change maintain the' transfer capability for the control power circuits. These items have the same qualifications as the original cable and circuit breaker. Therefore, this change does not increase
         ~t he probability or consequences of a malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). This change revises the DC power feed to the DG control panel to increase the voltage level at the panel and to provide assurance
         ' that the diesel generator has the necessary power to start and run when required. The DG is used
         .in response to certain accident scenarios, but is not credited as the initiator of any accident evaluated in the USAR.~ In the event of a Division 1 DG failure, the plant would rely on the separate and redundant Division 2 equipment for safe shutdown. This activity does not change the consequences of a failure of the DG. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. This design change does not alter the starting logic and operation of the diesel generator. Failure of the new relay is not a concern, because its output is only used for annunciation and not for operation of the DG or any other equipment; The wiring changes are not complex and the post modification testing confirms
         . that all fuses are connected to their appropriate source, The wiring changes do not cause any equipment to be operated i . s manner different from the original design. Therefore, this change does not create the possibility of an accident or malfunctiu of equipment important to safety of a         j
   ,     ' different type than previously evaluated. By revising the design so that the capability of the diesel
         . to start and run is maintained, this design change ensures that the operation of the plant as defined

Attachment 2           .
e                                ,                                                      to U-603276 Page 94 of 252
       'in the Technical Specifications is mamtained. Therefore, this activity does not reduce a margin of safety as' defined in the basis for any Technical Specification.
REPLACEMENT OF DIVISION III 480V - 120V' DISTRIBUTION TRANSFORMER 1 AP78E Activity Evaluated: Modification AP-035, Supp 1                                 Log Number: 98-034       -
Supplement 1 of Modification AP-035 replaces an existing 480-120 volt transformer, l AP78E, with a new 480-120 volt regulating transformer, I AP78ERT. This modification places the 120 volt auxiliary power and instrument power distribution systems in an analyzed configuration, providing voltage regulation to .120 volt loads. This change allows the low voltage power system .
to maintain an acceptable voltage range during worst case design basis conditions oflow or high voltage at the 4.16 kV level. The function of the new regulating transformer is to provide 120 volt alternating current (AC) power to the 120 volt loads connected to the Class IE Division III            l
      . Distribution Panel.: This modification evaluates the plant while operating in Modes 4 or 5, entry into Modes 1,2, or 3 is not covered by this modification.
Partial or complete failure of the transformer and loss of some or all of the loads powered from it is not an assuined accident initiator. The new regulating transformer meets the design, material           <
and construction standards used during original construction of the Division III onsite power system. The transformer is located in the same area of the plant and the design of the conduit and cable installation meets the same requirements for separation and independence as the original.
Installation and wiring of the transformer is in accordance with piant construction and maintenance procedures; Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the Updated Safety Analysis Repon (USAR). In the original design, the breakers on individual
: loads trip _ individually without any affect on the other secondary loads. However, if the current exceeds 150-200% of the new transformer rating, a short duration loss of power may occur until -
the fault is cleared from the circuit. Voltage then returns to normal for the uninvolved circuits.
Calculations show that these loads would be without power only for faults on circuits connected Lto t.he same phase as the fault. There are 3,4, and 3 circuits for phases A, B, and C respectively;
      ~t his limits the overall affect of a single occurrence to a small number of devices. Each of the type of components that 'could lose power for' a brief period do not perform a function immediately edtical to safety.- This short duration lossLof power has thus been determined not to be a critical factor in the' performance of the safety function of the connected loads. Therefore, this change
       . does not create the possibility of an accident or malfunction of eguipment important to safety of a
      . different type than previously evaluated. Technical Specification Bases 3.8.10, Distribution
       . Systems - Shutdown, requires that the distribution systems be operable in Modes 4 and 5.
Installation was performed within the Technical Specifications Limiting Conditions of Operation to maintain required suppon systems operable. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
l
                               -                                                                                    )

Attachment 2 to U-603276 Page 95 of 252 EXTENDED EVALUATION TEST OF DIVISION 2 NSPS INVERTER
           ' Document Evaluated: CPS Procedure 2813.01, R/l .                                Log Number: 98-035 Clinton Power Station (CPS) Procedure 2813.01, " Extended Evaluation Test of IC71-S001B Invertet," is a special test procedure for performing evaluation testing of Division 2 Nuclear System Protection System (NSPS) inverter. .This procedure helps identify why the inverter-
          - (IC71-S001B) transfers the Division 2 NSPS bus from the normal inverter power supply to the alternate. power supply for no identifiable reason. In order to perform this test, the Division 2 NSPS inverter needs to be INOPERABLE. While the test equipment is being installed as well as during Phase III testing, the inverter is " Tagged Out" and the NSPS bus is powered from the regulating transformer. The transformer breaker is protected by a special device that prevents manual actuation of the breaker. This breaker prevents the inadvenent loss of the Division 2 NSPS bus and shutdown cooling. Testing consists of three phases. During Phase I, the test equipment, inverter and plant equipment operate in a normal manner. A separate procedure and safety evaluation handle the Phase II portion. The inverter is isolated from the rest of the plant .
during Phase III. This test is performed in Mode 4 only..
The inverter is not being relied upon to perform any safety function. The addition of the test equipment does not disable the protective features of the breakers. This procedure does not prevent the NSPS bus from responding as it normally would. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Repon (USAR). Loss of the inverter causes the power to transfer from the DC to the alternate AC source. The test does not prevent this transfer."Also barriers are in place that prevent' loss of shutdown cooling that could' occur if the
          ' Division 2 NSPS bus were to be inadvenently lost. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. During performance of the test, the inverter operates in a normal manner. The test equipment is connected to the invener in a secure manner and has been chosen to prevent any impact on the inverter or the connected safety equipment.' Therefore, this change does not create the possibility of an accident or equipment malfunction of a different type. Technical Specification 3.8.8 requires only one inverter in Mode 4. During the testing, the Division 2 NSPS inverter is considered INOPERABLE, however, the Division 1 NSPS inverter operates in a normal nianner.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical
         ' Specification.-
TEMPORARY POWER TO RADWASTE BUILDING RISER B (0AP38E)
Activity Evaluated: TM 98-022                                                    IJog Number: 98-036-Temporary Modification'(TM) 98-022 provides temporary electrical power to Radwaste Riser B (electrical bus 0AP38E) from Unit Sub E (0AP45E3C) during the 4-kV breaker rework on bus 1 AP08E. The temporary power is necessary to maintain lighting in the 762' elevation of the Radwaste Building and outdoor security lighting. The buses affected by this temporary A                              _.a

                                                                                          . Attachment 2            ,
to U-603276 Page % of252 L modification are nonisafety related and are not required to furnish power under accident conditionsD This change does not. impact any design basis accidents evaluated in the Updated
        . Safety Analysis Report (USAR). All equipment that is affected by this modification are normally
: fed from the non-safety 4.16-kV Bus IB." This temporary modification replaces the normal
        . non-safety source with an alternate non-safety source 6.9-kV Bus l A. Since all loadings and R       'short-circuit contributions are within equipment and cable ratings, the equipment functions the same with either the 4.16-kV Bus lB supply or the temporary 6.9-kV Bus l A supply. These buses are neither assumed initiators, nor function in a mitigatory capacity during any previously
        > evaluated accident. Therefore, this change does not increase the probability or consequences of
        . an accident or malfunction of equipment important to safety previously evaluated in the USAR.
lAll electrical loadings are within equipment and cable ratings.f Breaker settings are adjusted to
        . provide overload protection as well as coordination with upstream breakers. An assessment demonstrates that the installatio_n is acceptable from a structural and seismic point ofviewi Also,-
: this activity' does not change the overall system performance < Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different
: type than previously evaluated in the USAR. This tempor. 3 modification does not change :
        . system response characteristics; nor does it cause system operation outside ofits design limits, operational transients in the system, or adverse system interaction'with other systems. Therefore,
         ' this activity does not reduce a margin of safety as defined in the basis for any Technical
        . Specification.
I EXTENDED EVALUATION TEST OF DIVISION 2 NSPS INVERTER Document Evaluated:.. CPS 2813.01 R/1, PAC 0086-98                             . Log Number: 98-037 and CPS 2813.01 DOC i R/1,
                                    -PAC 0087-98 Clinton Power Station (CPS) Procedure 2813.01, " Extended Evaluation Test of IC71-S001B Invener,".is a special test procedure for performing evaluation testing of the Division 2 Nuclear System Protection System (NSPS) invener. This procedure is intended to identify the cause of
        . the invener (1C71-S001B) transferring the Division 2 NSPS bus from the normal inverter power supply to the alternate power supply for no identifiatie reason. . In order to perform this test, the
         . Division 2 NSPS inverter will be declared INOPERABLEi Testing consists of three phases.
During Phase I, the test equipment, inverter and plant equipment operate in a normal manner. A separate procedure and safety evaluation handle the Phase II portion.' The inverter is isolated from the rest of the plant _during _ Phase III. While the test equipment is being installed as well as during Phase III testing, the inverter is " Tagged Out" and the NSPS bus is powered from the
: regulating transfo'rmer. The transformer breaker is protected by a special device that prevents
         - accidental manual actuation of the breaker. This action will help prevent the inadvertent loss of
        - the Division 2 NSPS bus and shutdown cooling. This test is performed in Mode 4 only.
The inverter is not relied upon to perform any safety function during Mode 4 operation. The addition of the test equipment does not disable the protective features of the bus and associated l

  'c         . .
     * '                                                                                                  Attachment 2 U                                                                                                         to U-603276 Page 97 of 252 electrical breakerse This procedure does not prevent the NSPS bus from responding as it normally would. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis l                    ' Report (USAR). ' Loss of the inverter causes the power to transfer from the DC source to the L,
alternate AC source. The test does not prevent this transfer. Also barriers are in place that prevent loss of shutdown. cooling that could occur if the Division 2 NSPS bus were to be inadvenently de-energized! Therefore, this change does not increase the probability or
: consequences of an accident previously evaluated in the USAR. During performance of the test, the inverter operates in a normal manner. The test equipment is connected to the inverter in a secure manner and has been chosen to prevent any impact on the inverter or the connected safety related equipment. Therefore, this change'does not create the possibility of an accident or Lequipment malfunction of a different type. Technical Specification 3.8.8 requires only one inverter in Mode 4. - During the testing, the Division 2 NSPS inverter is considered
                      ; INOPERABLE, however, the Division 1 NSPS inverter operates in a normal manner. Therefore, G           , this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
CONTINUOUS THERMAL OVERLOAD PROTECTION BYPASS FOR CROSS-TIE l-                      VALVES BETWEEN THE PLANT ' SERVICE WATER AND SHUTDOWN SERVICE WATER SYSTEMS p
c Activity Evaluated: ORM Change 19-5                                                 Log number: 98-038
                      . This activity revises' Attachment 3-5 to the Operational Requirements Manual (ORM) by
                      ~ designating on the component list that valves ISX014A, ISX014B, and ISX014C only have thermal overload protection bypassed in the closed direction.- This is in compliance with
                                                                            ~
Regulatory Guide (RG) 1.106: Updated Safety Analysis Report (USAR) 9.2.1.2.1.2 states that these valves are required to close and isolate the Plant Service Water (WS) system from the
                      ~ Shutdown Service Water (SX) system on a design basi.s Loss of Coolant Accident and on a Loss of Offsite Power. The ISX014A, B,'and C valves have thermal overload protection to protect the valve motor.' The thermal overload protection is byps sed to prevent inadvenent actuation of the th'ermal' overload protection and thus allow the valve.to complete its safety function. This thermal overload protection bypass is continuous and pan of the original licensing basis for the plant.
                      . This change corrects an inadvenent error in the equipment list and does not make any physical changes to.the valve actuator or any other component. - This change does not compromise the
                       ~ design, construction, or material standards of the valve. Also, this change does not have any        l
                      ~affect on the thermal overload protection or the bypass so as to change the component response or performance charact ristics for the valve. This activity does not change any parameters,
                      . environmental,' seismic, quality, or separation qualifications. Therefore, this change does not
: increase the probability or consequences of an accident or malfunction of equipment imponant to safety previously evaluated 'inl the USAR. This change does not introduce any component into the
[l L        4               plant, nor does it physically change any component, system or structure in the plant. The original   ,
L                     ; design and licensing basis required thermal overload protection bypasses only in the direction of h                        the' safety function, which is the closed direction for this valve. -Therefore, this change does not b
I
(
             )     )                           t

Attachment 2              ,
to U-603276 Page 98 of 252
     . create the possibility.of an accident or malfunction of equipment important to safety of a different type than previously evaluated. . The requirement that the thermal overload protection devices for the ISX014A, B, and C valves be continuously bypassed in the direction of the safety ftmetion was relocated from the Clinton Power Station (CPS) Technical Specifications t_o the ORM during the transition to the Improved Technical Specifications. The requirement in the ORM has not changed. The equipment list was inadvertently changed to show that these valves had continuous bypasses in both the open and closed direction.. This error was made during the transition of the
    ! equipment list from the CPS Standard Technical Specifications in Amendment 68 and the .
transition to the.ORM at the time of Amendment 95 to the CPS license. The Nuclear Regulatory
    . Commission (NRC) accepted these valves on the basis that the valves had continuous thermal
    . overload protection bypas'ses in accordance with RG 1.106. The thermal overload protection bypass in the direction of the. safety function has not changed, and the NRC acceptance is not J compromised. Therefore, this change does not compromise the acceptance criteria for these
     . valves, and as such, does not reduce a margin of safety as defined in the basis for any Technical Specification.
_   RELOCATION AND SEISMIC RESTRAINT OF CONTROL ROOM SCBA UNITS
    = Activity Evaluated: ECN 30778; USAR Change 8-228                               Log Number: 98-040 Engineering Change Notice (ECN) 30778 provides storage cabinets for the Main Control Room (MCR) self-contained breathing apparatus (SCBAs) and locates the units throughout the MCR envelope. Sections 6.4.4.2 and 2.2.3.1.3 of the Updated Safety Analysis Report (USAR) discuss the threat to MCR habitability due to the toxic gas event and toxic chemir.al release. The only credible event is the '' instantaneous release of ammonia due to failure of one or two of the 40 ton tank's at Terra International leading to a concentration level in excess of the MCR habitability limit at the MCR air intake." Probability of a tank failure was calculated as 0.000001 per year, and therefore,"is~ acceptably low enough to not be considered a design-basis event." However, the plant 'was designed to cope with a toxic gas event. Regulatory Guides 1.78 and 1.95 requires -
using a breathing apparatus within two minutes after an alarm. The SCBA storage cabinets are
: designed such that they will not interact with any safety related equipment. Relocation and seismic restraint of the SCBA units provide additional assurance that the operations staff can             !
safely perform accident or transient mitigating functions following a' toxic gas event. The relocation of the SCBA units allows operations staff to access the units much quicker than was             J
    . previously possible in their former location. Seismic design of the cabinet assures there will be no interaction with. safety related equipment. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety
    ; previously evaluated in the USARJ The only failure modes associated with the SCBA units are the potential interaction'with safety related systems and the potential of a missile generation due to the rapid de-pressurization of a SCBA cylinder.. The SCBA units are seismically rWrained in a storage cabinet and the~refore, are not capable of creating an3failure modes with safety related control room equipmenti The SCBA units are stored in hard plastic cases and the cases are stored in a seismically restrained storage cabinet. Nuclear Regulatory Commission Information 1
i l
1 3

r-c                                                                                      Attachment 2 to U-603276 Page 99 of 252 Notice 91-37 identified a potential missile hazard with compressed gas cylinders while performing normal maintenance and testing. Based on this notice, provisions have been specified in the modification for the SCBA cases to_ be removed from the MCR prior to performing any -                   l maintenance or functional testing, thus eliminating the possibility of missile pneration.                l Therefore, this activity does not create the possibility of an accident or malfunction of equipment l      important to safety of a different type than previously evaluated. There are no Technical              -
Specifications or Limiting Conditions for Operation associated with the location or use of the SCBA units. Therefore, this activity does not reduce a margin of safety as defined in the basis for I
any Technical Specification.
MODIFICATION OF CONTROL CIRCUITS FOR VALVES IG33-F001, -F004, -F101 AND -F102 :
           - Document Evaluated: Modification RT-039, Rev 0;                              ' Log Number: 98-041R/l USAR Change 8-120, and -                                                   ,
                                         - ORM Change 23                                                         ]
Nuclear Regulatory Commission Information Notice 92-18 identifies a scenario which can result
           . in valve damage due to hot shorts in the control wiring of Motor Operated Valves (MOV) in the
: event of a Main Control Room (MCR) fire. Modification RT-039 modifies the control circuitry for four MOVs (lG33-F001, IG33-F004, IG33-F101, and IG33-F102) and adds transfer / control switches for valve IG33-F004 to allow valve closure from outside the MCR. The control circuitry is rewired to place the open/close coils between valve limit / torque switch circuitry and the MCR circuitry. This prevents the bypassing of the limit / torque switches if there is a hot short
           ' in the MCR wiring and protects valve and interface piping from physical damage. Even though the open/close coils of the MOVs are powered by the AC source, the coils are disconnected from .
the grounded terminal of the c'ontrol power transformer. _ Caution statements are added to the respective Motor Control Centers (MCC) to provide a warning to plant personnel of the
            . ungrounded condition.
: These changes are made to prevent valve damage and/or breach of the pressure boundary in the evsnt 'of a fire in the MCR.. The capability to respond to design basis accidents has not been -
Ediminished? After the transfer opewtion, the automatic isolation signal and the automatic opening permissive for the IG33-F004 valve is not available. The transfer switch is only used during safe shutdown after a fire in the MCR and the valve is closed to maintain the Reactor Core Pressure Boundary (RCPB). Therefore, this modification does not increase the probability or consequences of an accident previously evaluated in the Updated Safety Analysis Report (USAR).
The new transfer / control switches are models already used for safety related application at Clinton Power Station in similar plant environments. - If the switches failed, the results are not any worse c       ~t han prior to the addition of the switch. The response of these valves to the operation of their respective control switches is exactly the same after the implementation of this modification as it      !
was prior to the modification. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the
            -- USAR. . The components installed at MCC 1 A3 provide capability for closure of valve l
i

Attachment 2            ^
to U-603276 Page 100 of 252
                                                   ~ IG33-F004 in the same manner as the MCR switch; there is no capability for opening the valve from the MCC controls. Therefore, this change does not create the possibility of an accident or
                                                   . equipment malfunction of a different type.: The bases of th'e Technical Specifications establish the minimum system requirements to assure.the validity of the analysis provided F. the USAR. The changes being made to the valves do not invalidate any of these bases, and t.5erefore, do not reduce a margin of safety as defined in the basis for any Technical Specification.                       -
EXTENDED EVALUATION TEST OF DIVISION 2 NSPS INVERTER - PHASE IV TESTING
                                                  ; Document Evaluated: CPS Procedure 2813.01, Rev.1                                Log Number: '98-042 PAC 0162-98
                                                  - This Safety Evaluation' covers Procedure Advance Change (PAC) 0162-98 to Clinton Power Station (CPS) Procedure 2813.01, Revision 1, " Extended Evaluation Test of IC71-S001B
                                                  - Inverter." PAC 0162-98 was prepared to perform Phase IV evaluation testing of Division 2 Nuclear System Protection System (NSPS) inverter IC71-S001B in its normal enclosed configuration while monitoring Division 2 NSPS Bus IC71-P001B loads. The test aids in identifying the cause of the Division 2 NSPS inverter transferring inadvertently from the normal inverter power supply to the alternate power supply.
>                                                    The monitoring does not block the automatic transfer of the Division 2 NSPS bus from the inverter to the Division 2 NSPS regulating transformer. The purpose of the test is to find the cause of the unexpected transfers. All the equipment fed from the Division 2 NSPS bus operate in
                                                   . a normal manner and are able to respond as described in the Updated Safety Analysis Report.
Therefore, this change does not increase or create the probability or consequences of an accident.
This PAC restores the inverter enclosure to permit normal operating temperatures within the
                                                    ' inverter. The test equipment is' passive, consisting of clamp on current detectors and high impedance voltage probes. The test equipment does not disable any of the inverter trips nor does           1 it prevent the inverter from transferring to the alternate source. Therefore, this change does not         j
                                                   . increase or create the probability or consequences of an equipment malfunction. Technical                  i Specification 3.8.8 requires only one inverter in Mode 4. During the testing, the Division 2 NSPS inverter is considered INOPERABLE; however, the Division 1 NSPS inverter operates in a normal manner. Therefore,' this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
                                                  ! FACILITY CHANGES IN THE RP OFFICE AREAS l
                                                   . Activity Evaluated; . CR l-97-02-203, USAR Change 8-071. ,                     Log Number: 98-043 and CIEL Items 90-05 and 90-06                                                    )
Configured item Exception List (CIEL) 90-05 removes shelves and overhead bookcases from the Radiation Protection (RP) Counting Room and converts the area into an office for the Supervisor Radiological Operations. CIEL 90-06 removes sliding windows above the counter on the east I
f
  . . . . . . . . ~ _ . - . . . . . . . . . . . .
                                                                                      --------------u---

;=                                                                                       Attachment 2 to U-603276 Page 101 of 252 y         ,
l        side of the RP Technician Oflice. These changes affect Updated Safety Analysis Report (USAR)          ,
;        Figures 1.2-17 and 12.3-38. Counting can be done more efficiently in the Radiological Control         !
I        Area (RCA) where a large majority of the samples are taken. The whole body counter was moved to the Emergency Operating Facility. Moving the whole body counter and performing the sample counting in the RCA only affects the locations where the tasks are perfonned. These are
       . configuration changes to the RP oflices only, they have no affect upon any accidents, accident mitigating equipment or equipment important to safety. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. These configuration changes are limited to facilities that have
       . no affect on the operation of the power producing equipment or even the operatic.nal                  l administration of the plant. These changes do not create the possibility of an accident or            )
malfunction of equipment important to safety of a different type than previously evaluated in the USAR. This activity does not degrade any safety limits, limiting conditions for cperation, or design limitations discussed in any Technical Specification. Therefore, this actLity does not reduce a margin of safety as defined in the basis for any Technical Specification.-
FACILITY CHANGES IN THE RADWASTE BUILDING EL. 762' OFFICES                                            ,
l Activity Evaluated: CR l-97-02-203, USAR Change 8-072, .                      . Log Number: 98-044    I and CIEL 90-07 Configured Item Exception List (CIEL) 90-07 removes a wall and the door on the south side of the Electrical Control and Instrumentation assistant superintendent's oflice. These changes affect
;        Updated Safety Analysis Report (USAR) Figures 1.2-6,12.3-13,17, FP-19a, and FP-19b.
CIEL 90-07 also allowed other changes, but are not mentioned in the USAR nor identified on any        )
Figures. This activity does not affect any of the accident analyses discussed in the USAR. This       !
activity involves configuration changes to the maintenance oflices only. It does not affect any accident mitigating equipment or equipment important to safety. The wall and door do not function in any capacity as barriers to the release of radiation during normal, abnormal, or          !
accident conditions.- Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
These configuration changes are limited to facilities that have no affect on the operation of the .   ;
f        power producing equipment or operational administration of the plant. Therefore, this activity l      Ldoes not create the possibility of an accident or malfunction of equipment important to safety of a l      ' different type t_han previously evaluated in the USAR. This activity does not degrade any .
Technical Specification safety limit, limiting condition for operation, or design limitation.
,?       Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
i

Attachment 2             .
to U-603276 Page 102 of 252 TEMPORARY PROCEDURE CHANGE TO TEST DIVISION 11 SHUNT TRIPS Activity Evaluated: TPD to CPS No. 9061.04' R/36                               Log Number: 98-045 This Temporary Procedure Deviation (TPD) to Clinton Power Station (CPS) Procedure 9061.04 disables various system components and manually injects a Division 2 containment isolation -
actuation signal in order to verify operability of various Division 2 components. This change also -
adds a step to check that the shunt trips reset or close prior to testing completion and to verify the
           ! shunt trip breakers trip as required, by initialing the steps. Without taking additional precautions, injecting a containment isolation signal would result in closure of IE12-F009, Residual Heat
           -Removal (RHR) Shutdown Cooling inboard suction valve. This TPD maintains this valve in the open position by opening the associated breaker with the valve in the open position. This action
       . ensures that RHR is not lost during the conduct of the surveillance.
2.
No system equipment is operated outside its design limits, allowed to cause adverse system
          ; interaction, increase the potential to drain the vessel, or induce a loss of RHR.' Due to the plant being'in cold shutdown during the performance of this TPD, the only previously evaluated accident that could be impacted is Failure of RHR Shutdown Cooling. Since the IE12-F009 valve is being deactivated in the open position, the probability of this event is not increased. The systems assumed to mitigate this event are not disabled by this TPD. The systems that are left in l-          operation are operated per their respective system operating procedure. The valves that are -
manually closed are valves that would have closed automatically when injecting the isolation
          ' signal. These valves / systems are disabled to prevent unwanted operation. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). No
          . equipment is being operated in a manner different than previously evaluated. None of the changes
           - degrade the performance of any system below the design basis. Therefore, this change does not create the~ possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. Technical Specification Limiting Condition for Operation 3.6.1.?
requires either IE12-F009 or IE12-F008, RHR Shutdown Cooling outboard suction valve, to be capable of automatic closure on low reactor water level. Due to having its power removed during
-           performance of this TPD, IE12-F009 is 'not capable of automatic closure on a low reactor water level signalc However, redundant isolation valve IE12-F008 will still isolate on a low reactor
          - water level signal from Division 1. This TPD does not change system response characteristics; nor does it cause system operation outside ofits design limits, operational transients in the system, or adverse system interaction with other systems. Therefore, this change does not reduce a
           -margin of safety as defined in the basis for any Technical Specification.                                  !
USAR CHANGE TO ALLOW OFF-SITE VENDOR PROCESSING OF WASTE Document Evaluated: USAR Change 8-117 Log Number: 98-046 This change revises Section 11.4.6 and Figure 11.4-1 of the Updated Safety Analysis Report
_(USAR) as follows: (1) correct the reference to drums to reflect the use oflarge volume

r
      .                                                                                             Attachment 2 to U-603276 Page 103 of 252 L
containers; (2) change "High-level waste" to " Higher activity Low-Level waste"; and (3) add "or licensed secondary waste processors for additional processing"in regards to shipment. The second change is necessary because federC regulations' define High-level waste as spent fuel and should not be addressed in Chapter 1I of the USAR. The third change reflects the technology advancements that allow off-site processors to more efliciently prepare waste for disposal. The first two changes are administrativein nature and have no impact on the probability of any accidents. For the third change, the waste is processed on site to a form suitable for transportation allowing Clinton Power Station (CPS) to ship the waste to a licensed off-site waste pro ,essor to achieve the best practical volume reduction. The accident most likely to be invoked
                  . would be a radioactive release due to liquid radwaste tank rupture. Sargent and Lundy evaluation, SLMI 24228, demonstrated that the small fraction (<10%) dose limits of Title 10 Pan 100 of the Code of Federal Regulations will not be exceeded from this event. The vendor's license covers the shipment from the off-site processor to final disposal. There are no interface connections to''any safety related systems or operations for waste processing on site; this change only deals with the disposition of the waste during the time between leaving CPS and final
                  < disposale Therefore, this change does not increase or create an accident or a malfunction of equipment important to safety. These changcs do not aff'ect any system involved with reactivity control or reactor safety. Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
_ DIVISION 2 NSPS DISTRIBUTION BUS LOAD MONITORING DURING PLANT
                   ' OPERATION-l Activity Evaluated: ' Temp orary Modification 98-025                           Log Number: 98-047 Temporary Mod' ification 98-025 provides means ofinstrumenting the Division 2 Nuclear System
,                  _ Protection Sy's tem (NSPS) Bus IC71-P001B loads during normal plant operation to determine
                   - their affect on the Division 2 NSPS invener, This includes installation of ring type current transformers (cts) at the NSPS load breakers, extension of test leads to the recorder through an l                    externally mounted test switch. While the temporary modification is being installed and removed, the Division 2 NSPS bus will be de-energized.~ . One Divisional inverter associated with the b                    Division 1 or 2 onsite Class IE uninterruritible AC bus electrical power distribution subsystem (s) p                    is required to be OPERABLE. The Division 1 invener is available during this time. Updated-f1          RL       Safety Analysis Report (USAR) Chapter 15_ evaluates Loss of Shutdown Cooling accident with respect to loss of all offsite AC power. The installation of non-safety related cts and terminal
   +        -
                   ~ block have been evaluated for structural integrity, fault propagation, and channel separation.
q               - With the temporary modification installed, Division 2 NSPS Bus IC71-P001B is powered in a
                    . normal manner through the inverter. The NSPS distribution bus loads are monitored during all
  ?                  Modes of plant operation. The inverter is operated in the same manner as it is no. mally operated h                    The temporary modification does not disable any of the inverter trips or shutdow:, features, nor l
                   - does it prevent'the inverter from transferring to the alternate source from the DC to AC buses.
All plant'equipsnent performs their intended function in a normal manner. Most monitoring equipment is non-intmsivef Monitoring equipment and leads actually interfacing with the system have been evaluated for seismic adequacy, channel separation, short circuit potential, and fault l
y

Attachment 2               ,
to U-603276 Page 104 of 252 propagation. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the USAR. The inverter can fait causing high voltage, low vohage, loss of sync or shorts. The resulting efv    ec ts of these failures is either a transfer to the alternate source or loss of the NSPS bus. The loss of NSPS bus result in the Loss of Shutdown Cooling, previously evaluated. Therefo:e, Jts activity does not create the possibility of an accident or malfunction of equipment important to safety of a           -
different type than previously evaluated. The Technical Specifications do not descrite details of NSPS Distribution bus circuit configuration. Technical Specification 3.8.8 and Bases 3.8.7 describes the requirements for preferred source of power for the uninterruptible AC buses. The inverters are required to be OPERABLE in Modes 1,2, and 3, but only one inverter is required to be OPERABLE in Modes 4 and 5. The Division 1 inverter is available during installation and removal of this temporary modification. Once installed, the invertei will again be fully Operable and satisfy all Technical Specification requirements. Therefore, this temporary modification does not reduce a margin of safety as defined in the basis for any Technical Specification.
MODIFICATION OF CONTROL CIRCUITS FOR INSTRUMENT AIR VALVES Document Evaluated: Modification IA-021,                                    Log Number: 98-049 R/l USAR Change 8-118, and ORM Change 23-2 Nuclear Regulatory Commission Information Notice 92-018 identifies a scenario which can result in valve damage due to hot shorts in the control wiring of Motor Operated Valves (MOV) in the event of a Main Control Room (MCR) fire. Modification IA-021 modifies the control circuitry for Instmment Air valves IIA 012A and llA013A and adds transfer / control switches for valve lIA012A to allow valve operation from outside the MCR. The control circuitry is rewired to place the open/close coils between valve limit / torque switch circuitry and the MCR circuitry. This prevents the bypassing of the limit / torque switches if there is a hot short in the MCR wiring and protects valve and interface piping from physical damage. Even though the open/close coils of the MOVs are powered by the AC source, the coils are disconnected from the grounded terminal of the control power transformer. Caution statements are added to the respective Motor Control Centers (MCC) to provide a warning to plant personnel of the ungrounded condition.
These changes are made to prevent valve damage and maintain backup air supply for the Division I Safety-Related Valves in the event of a fire in the MCR. The capability to respond to design basis accidents has not been diminished. After the transfer operation, the automatic opening response for the lIA012A valve is not available. The transfer switch is only used during safe shutdown after a fire in the MCR and it is not required to postulate a Design Basis Accident.              j Therefore, this modification does not increase the probability or consequences of an accident                    l previously evaluated in the Updated Safety Analysis Report (USAR). The new transfer / control switches are models already used for safety related application at Clinton Power Station in similar plant environments. If the switches failed, the results are not any worse than prior to the addition of the switch. The response of these valves to the operation of their respective control switches is exactly the same after the implementation of this modification as it was prior to the modification.
l l
l

p
  '                                                                                     Attachment 2 to U-603276 l
Page 105 of 252 l
l Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The components installed at MCC 1 A3 provide capability for closure of valve lIA012A in the same manner as the MCR switch. There are no mechanical or structural changes involved which introduce the possibility of a new physical failure. Therefore, this change does no.t create the possibility of an accident or equipment malfunction of a different type. The bases of the Technical Specifications establish the minimum system requirements to assure the validity of the analysis provided in the USAR. The changes being made to the valves do not invalidate any of these bases, and therefore, do not reduce a margin of. safety as defined in the basis for any Technical Specification.
l      TEMPORARY PROCEDURE CHANGE TO TEST DIVISION I SHUNT TRIPS                                               i Activity Evaluated: TPD to CPS Procedure 9061.04 R/36                            Log Number: 98-050 This Temporary Procedure Deviation (TPD) to Clinton Power Station (CPS) Procedure 9061.04 disables various system components and manually injects a Division 1 containment isolation actuation signal in order to verify operability of various Division I shunt trip circuits. This change also adds a step to check that the shunt trips reset or close prior to testing completion and to verify the shunt trip breakers trip as required, by initialing the steps. Without taking additional precautions, injecting a containment isolation signal would result in closure of IE12-F008, Residual Heat Removal (RHR) Shutdown Cooling outboard suction valve. This TPD maintains this valve in the open position by opening the associated breaker with the valve in the open position. This action ensures that RHR is not lost during the conduct of the smveillance.              !
No system equipment is operated outside its design limits, allowed to cause adverse system interaction, increase the potential to drain the vessel, or induce a loss of RHR Due to the plant being in cold shutdown during the performance of this TPD, the only previously evaluated               3 accident that could be impacted is Failure of Residual Heat Removal Shutdown Cooling. Since the IE12-F008 valve is being deactivated in the open position, the probability of this event is not     ,
      ' increased. The systerns assumed to mitigate this event are not disabled by this TPD. The systems that are left in operation are operated per their respective system operating procedure. The valves that are manually closed are valves that would have closed automatically when injecting the isolation signal. These valves / systems are disabled to prevent unwanted operation. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report               {'
      - (USAR). No equipment is being operated in a manner different than previously evaluated. None of the changes degrade the performance of any system below the design basis. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to         j safety of a different type than previously evaluated. Technical Specification Limiting Condition       l for Operation 3.6.1.3 requires either IE12-F008 or IE12-F009, RHR Shutdown Cooling inboard             i suction valve, to be capable of automatic closure on low reactor water levelc Due to having its        l power removed during performance of this TPD, IE12-F008 is not capable of automatic closure on a low reactor water level signal. However, the redundant isolation valve IE12-F009 will still isolate on a low reactor water level signal from Division II. This TPD does not change system 1
s l

v Attachment 2             .
to U-603276 Page 106 of 252 response characteristics; nor does it cause system operation outside ofits design limits, operational transients in the system, or adverse system interaction with other systems. Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical
      .' Specification.
      ~ RTD INSTALLATION ON RHR HEAT EXCHANGER FOR GL 89-13 TESTING                                                -
Activity Evaluated:: Temporary Modification 98-031                             Log Number: 98-051 Temporary Modification 98-031 installs test equipment to support testing of Residual Heat Removal (RHR) "A" Heat Exchanger (HX). Resistance Temperature Detectors (RTDs) are
      ^ being installed adjacent to the process piping for the purpose of verifying the capability of the heat exchanger to perform its design basi; function prior to mode change. The Shutdown Service r
      ~ Water (SX) system is not an assumed initiator of any accident. Since the RTD to pipe wall interface is metal to metal, or is made with an approved tape, no material incompatibility exists between the RTDs and the RHR and SX piping. Based on the low weight of the RTDs, there are no credible seismic interactions with the RHR and SX piping and supports.' The signal voltage
      -.and current used to develop a signal by the RTDs and transmitted by the wires is low, which means that the wires will not generate any heating. Potential combustible loading from the equipment and leads has been evaluated. The RTDs are passive and can not actively contribute to any system change in performance or operating characteristic. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment previously _
evaluated in the Updated Safety Analysis Report (USAR). The function and design basis of the RHR and SX systems are not altered by this temporary modification. Also, this activity does not
      . adversely affect the RHR "A" HX reliability and functionality. Therefo're, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The Technical Specifications do not require the performance of Generic Letter 89-13 testing; however, the. testing'does verify' operability of the RHR "A" HX.             4 RHR operability assumes that the design heat removal capacity of the RHR "A" HX exists.                     !
Since, the function and design basis of the RHR and SX systems are not altered and the RHR "A"             {
HX reliability and functionality are maintained, this activity does not reduce a margin of safety as
        . defined in the basis for any Technical Specification.
        ' RETIRE DRYWELL FLOOR DRAIN WEIR BOX FLOW LOOP AND RAISE WEIR BOX WATER LEVEL
         -Activity Evaluated: ECN 30300 and USAR Change 8-190 -                         Log Number: 98-052 Engineering Change Notice'(ECN) 30300 removes the function of the Drywell Floor Drain Weir Box Flow Loop, IE31-N580, by retiring it in place and raises the normal water level in the weir box. The Drywell Floor Drain Weir Box Flow Loop has been highly inaccurate and unreliable due to constant fouling of the v-notch weir plate. This change retires the loop in place by removing or lifting and taping the appropriate wiring in accordance with plant procedures. Power to the level transmitter will be eliminated and the affected computer points and annunciators will b
I

   .                                                                                   Attachmcnt 2 to U-603276
 -                                                                                     Page 107 of 252 no longer be used. The v-notch weir plate will be replaced by a solid plate to raise the normal water level in the weir box.
Updated Safety Am.ysis Report (USAR) Section 9.3.3.3 states: "The equipment and floor drain system is not required to assure either of the following conditions: a. the integrity of the reactor coolant pressure boundary, or b. the capability to shut down the reactor and maintain it in safe shutdown condition." The flow loop is a passive system for indication and alarm only. There are no automatic accident mitigating functions or control logic associated with the flow loop. This flow loop is a redundant system to the two existing reliable sump monitoring systems. These two systems are in place to separately monitor abnormal leakage to the sump and provide the function lost due to weir box fouling. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USA.R. The system important to safety that is impacted by the Drywell Floor Drain Weir Box Flow Loop is the Reactor Coolant Pressure Boundary (RCPB). The flow loop is classified as not related to safety and uses non-essential 120-VAC instrument power. This system provides qualitative information to the operator and was not considered reliable. The unidentified leakage going to the Drywell Floor Drain Sump is detected by two other monitoring systems which will trip and activate an alarm in the control room, as an early indicator of RCPB leakage. These are the primary detection methods for small unidentified leaks within the drywell. Due to the continued reliance on these leak detection loops, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The retired equipment will be de-energized, so the probable failure would be that of the two other leak detection loops. If the Floor Drain Sump monitoring system becomes inoperable, a 30 day Limiting Condition for Operation (LCO) is entered per Technical Specification LCO 3.4.7. Retiring the weir box loop leaves suflicient leak detection so no new accidents of a different type are postulated. The new solid plate will be attached in the same manner as the old v-notch plate and therefore have the same failure modes and consequences. Abandoning the associated flow loop instrumentation removes erroneous and unreliable information from the operators decision making process.
Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The requirements of the Technical Specification and Operating Requirements Manual are not impacted by this change.
Technical Specification Bases Change number BE-97-018 removed the non-safety related system from the Technical Specification Bases. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Spr :ification.
MODIFICATION OF CONTROL CIRCUlTS FOR FEEDWATER SHUTOFF VALVES Documett Evaluated: Modification FW-040,                                  Log Number: 98-054 R/l US AR Change 8-1M, and ORM Change 25-2 Nuclear Regulatory Commission Information Notice 92-18 identifies a scenario which can result in valve damage due to hot shorts in the control wiring of Motor Operated Valves (MOV) in the esent of a Main Control Room (MCR) fire. Modification FW-040 modiGes the control circuitry l

Attachment 2                               ,
to U-603276 Page 108 of 252 and adds transfer / control switches for Feedwater shutofTvalves IB21-F065 A and 1821-F065B to j      allow valve operation 1:am outside the MCR. The control circuitry is rewired to place the
(      open/close coils between valve limit / torque switch circuitry and the MCR circuitry. This prevents the bypassing of the limit / torque switches if there is a hot short in the MCR wiring and protects valve body and components, actuator, and motor from physical damage. Even though the open/close coils of the MOVs are powered by the AC source, the coils are disconnected from the                                                                                                                                                          -
grounded terminal of the control power transformer. Caution statements are added to the respective Motor Control Centers (MCC) to provide a warning to plant personnel of the ungrounded condition.
Operation of the transfer switch will annunciate in the MCR. Due to both handswitches at the MCC having to be utilized for this action to occur, this could not happen inadvertently.
Therefore, this change does not increase the probability ofinitiating Loss-of-Coolant Accident, High/ Low Energy Line Break, Loss of Feedwater Flow, Loss of Offsite Power or other accidents                                                                                                                                                             l previously evaluated in the Updated Safety Analysis Report (USAR). The new transfer / control                                                                                                                                                             l switches are models already used for safety related application at Clinton Power Station in similar                                                                                                                                                       l plant environments. If the switches failed, the results are not any worse than prior to the addition of the switch. The response of these valves to the operation of their respective control switches is exactly the same after the implementation of this modification as it was prior to the modification.
Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The revised control circuitry                                                                                                                                                             j maintains the same physical and electrical separation of divisional power sources as the original                                                                                                                                                         !
valve control circuitry. There are no mechanical or structural changes involved which preclude                                                                                                                                                            j the possibility of a new physical failure. Therefore, this change does not create the possibility of l
an accident or equipment malfunction of a different type. The bases of the Technical
{
Specifications establish the minimum system requirements to assure the validity of the analysis provided in the USAR. The changes being made to the valves do not invalidate any of these bases, and therefore, do not reduce a margin of safety as defined in the basis for any Technical                                                                                                                                                          !
Specifiv ation.
MODIFICATION OF CONTROL CIRCUITS FOR MAIN STEAM LINE OUTBOARD DRAIN ISOLATION VALVE Document Evaluated: Modification MS-044,                                                                                                                                                                Log Number: 98-055 USAR Change 8-133, and ORM Change 23-3 Nuclear Regulatory Commission Information Notice 92-18 identified a scenario which could result in valve damage due to hot shorts in the control wiring of Motor Operated Valves (MOV) in the event of a control room fire. Modification MS-044 modifies the control circuitry and adds transfer / control switches for Main Steam Line Outboard Drain Isolation valve IB21-F019 to allow valve operation from outside the Main Control Room (MCR). The control circuitry is rewired to place the open/close coils between valve limit / torque switch circuitry and the control room circuitry. This prevents the bypassing of the limit / torque switch if there is a hot short in the

a'                                                                                                             Attaciunent 2 to U-603276 Page 109 of 252 MCR wiring and protects valve and interface piping from physical damage. Even though the open/close coils of the MOVs are powered by the AC source, the coils are disconnected from the grounded terminal of the control power transformer. Caution statements are added to the respective Motor Control Centers :(MCC) to provide a warning to plant personnel of the l ungrounded condition.
                      , The safety related function of the Main Steam Line Outboard Drain Isolation Valve is to maintain the reactor coolant pressure boundary (RCPB) integrity. The changes to the circuitry prevent valve damage and /or breach of the pressure boundary in the event of a fire in the MCR. The .
transfer switch is only used during safe shutdown aner evacuation of the MCR and the valve is
                        - closed to maintain the RCPB Therefore, this change does not increase the probability of initiating Loss of Coolant Accident, High/ Low Energy Line Break, Loss of Feedwater Flow, Loss of Offsite Power or other accidents previously evaluated in the Updated Safety Analysis Report
                          '(USAR). The new transfer / control switches are models already used for safety related application
                        ^ at Clinton Power Station in similar plant ' environments. If the switches failed, the results are not any worse than prior to the addition of the switch. The response of these valves to the operation of their respective. control switches is exactly the same after the implementation of this l-                       -modification as it was prior to the modification.' Therefore, this change does not increase the
!                          probability or consequences of a malfunction cf equipment important to safety previously
                         / evaluated in the US AR. - The revised control circuitry maintains the same physical and electrice.1 separation of divisional power sources as the original valve control circuitry. There are no mechanical or structural changes involved which preclude.the possibility of a new physical failure.
L                       ' Therefore, this change 'does not create the possibility of an accident or equipment malfunction ~of a
!                       : different type. The bases of the Technical Specifications establish the minimum system l
                          ' requirements to assure the validity of the analysis provided in the USAR. The changes being
!                           made to the valves do not invalidate any of these bases, and therefore, do not reduce a margin of l                           safety as defined in the basis for any Technical Specification <
l
                            '6900 BUS 1 A (1 APO4E) OUTAGE ~
Activity Evaluated [ CPS Procedure 3514.01C001 Rev.1                                     Log Number: 98-056 Revision 1 to Clinton Power Station (CPS) Procedure 3514.01C001 provides instmetions,                       l materials list, and references for installation and removal of Temporary Modification 95-006. The           i instmctions were screened for safety implications when first installed. Those conclusions are still         l
[ valid, in that'a temporary, alternate source of power to Service Building Motor Control Center (MCC)"A" does not impact the manner in which any equipment is operated as described in any -
procedure discussed in the Updsted Safety Analysis Report (USAR). However, the original screening failed to identify the impact of the Temporary Modification on USAR Figure 8.3-1,
                                          ~
l which shows the normal polwer supply to the MCC. Thus, the original screening failed to identify            l the change to the facility as described in the US AR. This Safety Evaluation documents the acceptability of the Temporary Modification. This activity substitutes one non-safety related power source for another non-safety related power source. There are no loads associated with               !
Service Building MCC A or the 12kV loop which directly provide a safety function or support the            i i
n u                                 o

Attachment 3          ,
to U-603276 Page 110 of 252 operation of equipment important to safety, All components continue to operate within design
           ' parameters. Therefore, this activity does not increase the probability or consequences of an
          - accident or malfunctio_n of equipment important to safety previously evaluated in the USAR. No installed equipment is being operated in a new or different manner. This activity does not alter the
       ~'
           ~ parameters within which the plant is normally operated or the setpoints that initiate protective or
..        ' mitigative actions Therefore,'this activity does not create the possibility of an accident or             -
malfunction of equipment important to safety of a different type than previously evaluated in the -
USAR. This activity does not affect any Technical Specification safety limit, limiting safety system setting, or limiting condition for operation. Therefore, this activity does not reduce a
                                                      ~
margin of safety as defined in the basis for any Technical Specification.
DOCUMENTATION OF CONFORMANCE TO APPENDIX R, SECTION II.C
          , Document Evaluated; USAR Change 8-134                                             Log Number: 98-057
:         - Title 10, Part 50, Appendix R, Section II.C of the Code of Federal Regulations (CFR) addresses requirements for certain fire prevention features, for example: transient hazards, detection cixtinguishers, hose stations, barriers, suppression, fire brigade, personnel qualifications, and
     ,    - surveillance. This change revises Updated Safety Analysis Report (USAR) Section 9.5.1 and
                                            ~
Appendix E Sections 1.0,3.0, and 4.0 to comply with Appendix R. The following words are being added to USAR paragraph 9.5.1.b: "where applicable; fire prevention features are provided based on the results of the fire hazards analysis as provided in the CPS-FPER. Deviations for not providing required fire prevention features in areas that contain or present a fire hazard to
          . stmetures, systems, or components are provided in the CPS-SSA.". This change documents the Clinton Power Station (CPS) Fire Protection Program compliance with Section II.C of 10CFR50
            . Appendix R as currently reflected elsewhere in the US AR, and therefore does ~not affect identified
          . failure modes, standards, or system performance. Compliance and any deviation from compliance that would affect a postulated fire has been documented in Appendix E and F of the USAR,'and previously approved by the Nuclear Regulatory Commission in the Safety Evaluation Report (SER) and SER Supplements 1,3, 5, and 6. Therefore, this change does not increase or create the probability or consequences of an accident. This change does not reflect any change to the availability, performance, testability, or reliability of the Fire Protections System and other plant hardware, systeras, or procedures. Therefore, this change does not increase or create the
          - probability or consequences of an equipmert malfunction. All changes to acceptance values or
          ? design limitations to the CPS Fire Protection System have been previously approved by the NRC in the SER and SER Supplements. Therefore, this actiity does not reduce a margin of sefety as defined in'the basis for any Technical Specification.
ADDITION OF SLUICE GATE 2SX02S TO P&ID AND USAR FIGURE 9.5-1 SHEET 1 Activity ENluated: ECN 30814'and USAR Change 8-136                                Log Number: 98-058 Due to pop logs being in~ stalled in the Unit 2 traveling screen tunnels of the Circulating Water Screenhouse (CWS), fire pump B'(OFP0lPB) intake from the lake is only through sluice gate I-

.                                                                                    Attachment 2 to U-603276 Page 111 of 252 2SX02S which shall be locked open. Engineering Change Notice (ECN) 30814 adds sluice gate 2SX02S, detailed in the locked open position, to Piping and Instmmentation Diagiam M05-1039 Sheet I and Updated Safety Analysis Report (USAR) Figure 9.5-1 Sheet 1. Since the fire suppression system is not an accident initiating system, the probability of a fire occurring is not increased through changing the Division 11 fire pump suction path. The sluice gate is located on the Unit 2 side of the CWS which does not interface with the Unit I side of the CWS. Since the suctions for all three divisions of Shutdown Service Water (SX) pumps are located in the Unit 1 CWS, this activity has no affect on the SX system. This change does not add, delete or modify any component that would prevent the Fire Protection System from performing its function. The Fire Suppression System is designed such that normal or inadvertent system operation will not induce failure of safe shutdown systems. This activity does not affect the way the system is operated or the performance or design of the Fire Protection System. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment impmtant to safety previously evaluated in the USAR. This activity is limited to the Fire Protection System, but has no affect on the operation or function of the Fire Protection System and does not introduce any new or different equipment to the system. Therefore, this activity does not create an accident or malfunction of equipment important to safety of a different type than previously evaluated. The Technical Specifications do not address the Fire Suppression System. The Technical Specifications state that the station may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. This activity satisfies that criteria. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
CLARIFICATION OF BASES FOR TS 3.5.2, REQUIRED ACTION D3 Activity Evaluated: TS Bases Change BL-98-001                                    Log Number: 98-059 Technical Specification (TS) Bases for TS 3.5.2, Required Action D.3 provides additional detail concerning what is required to " restore isolation capability in each required secondary containment and secondary containment bypass penetration flow path that is not isolated." The Bases indicate that isolation capability can be effected by having sufficient automatic isolation   )
capability or by implementing adequate administrative controls to ensure suflicient manual isolation capability. The purpose of Technical Specification Bases change is to amplify the existing guidance as to what constitutes acceptable " administrative controls." This activity does not aher the design, material, or construction standards applicable to any plant Structure, System, or Component (SSC). This activity does not invoke any new testing or operational modes that         i could cause any plant systems or components to be operated outside of their design or testing limits. The qualifications, design functions, and performance criteria of all plant SSCs remain the same. The safety system ofimportance relative to the TS Bases revision is the secondary containment system. This activity does not impair or alter the availability of the secondary        l containment system. The change does not degrade performance in any manner and no systems including secondary containment will be impacted such that performance might degrade to less than the design capability. Secondary containment will operate as described in the Updated

m
                                             +
f:                                                  Attachment 2           ,
c                                              to U-603276 Page 112 of 252 Safety Analysis Report (USAR) accident analyses. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety
  ,             previously evaluated in the USAR[The bounding accident scenario is the Loss of Coolant Accident and there are no new or different accident scenarios made possible due to this change.
Since there will be no design or operational limits challenged or exceeded as a result of this
             , change, and since all equipment qualifications will remain unaltered, no new failure modes or             -
accident initiators will be introduced by the change. There is no equipment that will be realigned 4 -
or relocated to a new environment such that it will be subjected to. a previously unanalyzed l environmental condition as a result of this change. Therefore, this activity does not create the-possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The subject change only affects the TS Bases description as to what
         . l constitutes proper " administrative controls" when meeting TS 3.5.2, Required Action D.3.'
Implementation of this change does not impact the fuel cladding, reactor coolant system, or
              . primary containment; therefore, this activity'does not reduce a margin of safety as defined in the-
             - basis forany Technical Specification.
REDUCING S' RM C NOISE BY ADDING FERRITE BEADS AND GROUNDING PREAMPLIFIER CASE -
Document Evaluated: ECN 30831.                                                   L'og Number: 98-% 0 7         . Source Range. Monitor _ (SRM) channel C is experiencing increased noise levels on the input signal; Updated Safety Analysis Report (USAR) Section 7.1.2.1.4.1.l(1) states" Neutron sources and neutron detectors together shall result in a signal-to . noise ratio of at least 2:1. " Trouble shooting performed by the system engineer and an off-site noise reduction specialist determined that installation of ferrite beads and grounding the preamplifier chassis increases the signal to noise ratio by a factor of approximately 4. Engineering Change Notice (ECN) 30831 makes those changes.'
The fundamental function and objective of the system is not affected by this activity and the' activitywas analyzed not to have an adverse affect on any other system. - No credible failure mechanism exists that would allow the ferrite beads to cause an initiating event for control rod accident-or fuel handling accident. The beads are passive components and are not intrusive. The installation of the ferrite beads and grounding of the preamplifier are seismically analyzed.
Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. SRMs have no safety function and are not required to function during any design basis accident. The ferrite beads are banded to the external surface of Neutron Response system cabling, are'non-intrusive, and only affect the common mode high frequency noise.1 Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The failure modes'which could be introduced by this chanbe are failure of a ferrite bead, seismic interaction, the burning of a bead jacket, and the possibility of a Direct Current (DC) ground loop being created. Since the ferrite beads are not system intrusive, have no moving parts, do not require any external power, and are seismically ~ supported, the failure of a bead is not credible. Since the bead supports are

p
[         .
                                                              ~
    '.,                                                                                          Attachment 2 to U-603276 Page 113 of 252
                 ~ designed to be seismic, no failure of the suppons or the connected beads due to a seismic event -
need be postulated. Since the energy level of the signal is so low, the beads can not be the initiator of a fire? A'second ground path is created which is' acceptable for high frequency
(             - applications such as the SRMs. Therefore, this change does not create the possibility of an accident or equipment malfunction of a different type, Since the modification enhances the lP
              ; operation of SRM channel C and has no adverse impact on any other system, this activity does -           .
not reduce a margin of safety as' defined in the basis for any Technical Specification.
RAT AND ERAT FOUNDATIONS, GROUNDING AND BACKFILL 1
              . Activity Evaluated: ECNs 30815 and 30816                                         Log Number: 98- % 2          i 1                                      and USAR Change. 8-131                                                         ]
                                                                ~
Engineering Change Notices (ECNs) 30815 and 30816 provide minor changes to foundations,                   3 and material used for grounding and backfill in support of modifications to the Reserve Auxiliary         {
               ' Transformer (RAT) Static VAR Compensator (SVC) and Emergency Reserve Auxiliary .                           l Transformer (ERAT) SVC. This activity makes minor changes to the materials and
              ~ configurations previously specified to support fabrication of the foundations (not the foundation l
configuration or materials itself). This activity also changes the grounding grid cable size from         l 500 MCM to 400 MCM due to material availability. The backfill material changes from Type'B to CA-6. The CA-6 material serves to raise the elevation around the ERAT and RAT SVC areas                J and to provide freeze protection for piping below grade. The use of the different fill material is L limited to non-safety areas or above the frost linejust as Type B material was used. The changes-
                 'of mat'erial have been verified ~as equivalent. ' The change in the grounding grid cable size allows l       ,
use of 400.MCM cable in place of 500 MCM cable. Calculations confirm the capability of the smaller cable to function during the worst case fault conditions. The connection to the existing
                ; station grounding remains the same. Therefore, this change does not increase the probability or           i consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Repon (USAR).- Calculations for grounding and foundations                 ;
verify the equipment meet their design function and'the comparison of Type B material to CA-6            i finds minimal differences in strength and density. Because the alternative materials and designs i                  for the non-safety related foundations, backfill and grounding have been determined to be l              ~ equivalent to originals, the changes do not create the possibility of an accident or malfunction of 3
equipment important to safety of a different type than previously evaluated in the USAR. The'            !
r                  only Technical Specifications involv'mg the SSCs are those for Offsite and Onsite Power Systems.         -
3 The changes have no affect on any criteria or acceptance values for the Offsite Power System.
  =
Therefore, this activity does not reduce a margir of safety as defined in the basis for any Technical    f l              . Specification.
l'.
l l.
1 l

l
{
Attachment 2           ,
to U-603276 Page 114 of 252 ADD INTERLOCKS TO RH VALVES, ADMINISTRATIVE CONTROLS ASSOCIATED                                              ,
WITH " HOT SHORT" ISSUES                                                                                    l Document Evaluated: Modification RH-048 and                                      Log Number: 98-065 USAR Change 8-135 The first change of Modification RH-048 consists of providing interlocks between the Residual Heat Removal (101) system full flow test valve lE12-F024A, the RH suppression pool suction valve IE12-F004A, and the RH shutdown cooling suction valve IE12-F006A. The purpose of these interlocks is to prevent 1E12-F024A and/or IE12-F004A valves from opening if the IE12-F006A valve is not fully closed. This change removes the relays installed under Modification RHF011 and replaces the relays with existing limit switch contacts of the motor-operated IE12-F006A valve to enforce the interlock. The second change involves a design change to the control wiring circuits to prevent damage to IE12-F006A(B) valves due to " hot short" fires in the Main Control Room (MCR) as addressed in Nuclear Regulatory Commission Information Notice 92-18. The control circuitry is rewired to place the open/close contactor coils between the limit / torque switches and the remote handswitch. Even though the open/close coils of the Motor Operated Valves are powered by the AC source, the coils are disconnected from the grounded terminal of the control power transformer. Caution statements are added to the respective Motor Control Centers to provide a warning to plant personnel of the ungrounded condition. The third change involves imposing administrative requirements for valve IE12-F006B. During normal plant operations, the breaker is open to prevent inadvertent opening of the valve as a result of a " hot short" These alterations do not alter the function or interaction of the valves. Spurious operation of the valves as a result of a fire-induced short in the MCR is still a possibility; however, the valves are protected from physical damage. The alterations are internal to the control circuit for each of the valves and do not directly or indirectly affect the operation of any other component or equipment.
The changes to the interlocks affected by this change are not initiators for any design basis accidents evaluated in the Updated Safety Analysis Report (USAR). Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR.
The wiring changes involve physical relocation of the existing open/close coils between the limit / torque switches and the respective remote control logic for the valve. The operation logic and function of the valve have not been changed. Also, the limit switch in series with the torque switch is being eliminated. Originally, the limit switch was connected to prevent cycling of the valve due to relaxation of the torque switch at the end of the closing stroke. However, since the           l valve has an actuator with the locking gear capability, relaxation of the torque switch does not occur. Therefore, these changes do not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The interlocks provide a
{
more positive means of controlling shutdown cooling operations and reduce the probability of an             !
inadvenent draindown event. The action of placing suppression pool cooling into service requires            l operator actions to perform. However, the operator has sufTicient time to identify that a failure has occurred, and the failure is bounded by other failures. Therefore, this change does not create the possibility of an accident or equipment malfunction of a different type. The Technical a

  .                                                                                 Attachment 2 to U-603276 Page 115 of 252 Specifications for RH shutdown cooling and suppression pool cooling are not affected by this change; therefore, this activity does not reduce a margin of safety as defined in the basis for any  ;
Technical Specification.                                                                             l l
CHANGES TO SUBSTITUTE ADS (IN CONJUNCTION WITH LOW-PRESSURE ECCS)
FOR RCIC IN THE ACCIDENT ANALYSIS FOR A FEEDWATER LINE BREAK OUTSIDE CONTAINMENT Activity Evaluated: USAR Change 8-213                                   Log Number: 98-066 R/l This change revises the Updated Safety Analysis Report (USAR) to substitute the Automatic Depressurization System (ADS), in conjunction with low-pressure Emergency Core Cooling System (ECCS), for the Reactor Core Isolation Cooling (RCIC) System during a postulated Feedwater (FW) Line Break, within the Auxiliary Building Steam Tunnel, that renders RCIC not available due to room flooding. This activity does not involve any physical changes to the feedwater system, and as such, does not affect any failure modes that can act as an initiating event for the FW Line Break Outside Containment. This change does not alter any FW system                  )
operating characteristics nor the design, material, or construction of the FW piping. The revised analysis concludes that: if High Pressure Core Spray fails, ADS depressurizes the reactor vessel; reactor vessel water level decreases and the fuel is panially uncovered; low pressure ECCS begins injection to reflood the core; the event is terminated without fuel damage; and radiological consequences are not changed. This activity has no impact on the RCIC design basis for availability, performance, or any other accident analysis assumptions for the control rod drop event. Following a FW Line Break Outside Containment, the assumption that the FW check              j valves instantly isolate the reactor from the break, thus limiting the amount of reactor coolant released through the break remains unchanged. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The consequential failure of RCIC during a postulated i      feedwater line break outside containment would initiate a sequence of post-Loss of Coolant Accident mitigating events that are different than those presently assumed in the accident analysis l      for this event. However, this new scenario for mitigating the feedwater line break is no different than that previously evaluated for the main steam line break outside containment. The ADS, as a system in ECCS, is fully qualified for this event. Therefore, this change does not create the possibility of an accident or malftmetion of equipment important to safety of a different type than l      previously evaluated in the USAR. The substitution of ADS for RCIC does not affect RCIC or l      ECCS availability or degrade RCIC or ECCS performance for any design basis events for which l      they are required to function. Therefore, this activity does not reduce a margin of safety as
!      defmed in the basis for any Technical Specification.

I Attachment 2              ,
i to U-603276                 I Page 116 of 252 l
USAR CHANGE TO PROHIBIT THE USE OF THE OG COOLER DEFROST HEATER                                           '
Document Evaluated: USAR Change 8-196                                         Log Number: 98-068          )
I This change prohibits the use of the oft-Gas (OG) cooler defrost heater and specifies alternative methods to defrost the OG cooler. Restricting the use of the OG cooler defrost heater prevents the possibility of causing a fire in the OG charcoal beds. The OG vault refrigeration units (VO)          l can be shut off and the warm OG flow can defrost the cooler itself, or the VO system can be placed in the vault heating mode of operation to speed up the defrost. These changes impact Updated Safety Analysis Report (USAR) Section 9.4.8.2.g and Table 11.3-5. Restriction of off-gas flow is the only credible failure mode associated with this change that could cause an evaluated accident (e.g., Loss of Condenser Vacuum). The icing up of the cooler could reduce OG flow rate and could reduce the condenser vacuum. However, in order for the OG cooler to ice-up, a failure of the OG desiccant beds and regeneration system would have to occur.
Additionally, the gradual ice build-up would allow operators time to correct the situation and/or bypass the cooler for a short time. Also, the cooler can still be de-iced without the use of the defrost heater if the OG flow rate is afTected. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The OG system is a Gaseous Waste Management System and is a system important to safety. The gas cooler malfunction discussed in the USAR are for moisture freezeup due to failure of the OG desiccant bed function. The probability of this happening is not afTected by this change. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the US AR. The only conceivable accident or equipment malfunction that this change could affect is the flow of off gas due to icing up of the OG cooler which is already evaluated in the USAR, therefore, this change does not create an accident or equipment malfunction of a difTerent type. The Technical Specifications do not address the OG system; therefore, this activity does not reduce a margin of safety as defmed in the basis of any Technical Specification.
DRYWELL FLOOR DRAIN SUMP PUMPS AtR BINDING Document Evaluated: ECN 30840, USAR Change 8-185                              Log Number: 98-069 and Bases Change Request BE-98-016 Engineering Change Notice (ECN) 30840 corrects the problem associated with the Drywell Floor Drain Sump Pumps sporadic cavitation due to air binding. Plant Modification LD-028 established a low water level setting below the sump inlet pipe in order to maintain a linear level increment for leak rate determination. In order to resolve the air binding problem with the sump pumps, this change raises the sump operating water level above the inlet pipe. This maintains the pumps' inlet pipe full. The sump level rate of change calculation is not influenced by the non-linear characteristic of the inlet pipe volume. This change also replaces the existing mechanical alternator by transferring its functions to the Programmable Logic Controller installed in the Main Control Room (MCR).

p 3                                                                                           Attachment 2
;                                                                                               to U-603276
(,                                                                                              Page 117 of 252 lne Drywell floor' drain leak detection system does not perforni any active control functions, and it is not capable ofinitiating any accident previously evaluated in the Updated Safety Analysis
             ' Report (USAR). It only provides early indication of Reactor Coolant Pressure Boundary leakage.
Since this change does not invo' lve a change to the design intent of the system, this change does
_ not increase the probability or consequences of an accident previously evaluated in the USAR.
The purpose of this modification is to increase system reliability and maintainability by replacing   -
unreliable components which can not be serviced during plant operation. Therefore, this i
modification provides the same level of protection as the original design and continues to meet the
;               requirements of Regulatory Guide 1.45. 'As !.tated in the USAR, this system is not a safety related l          ;    system. Therefore, this change does not inc' ease the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This modification does not involve a change to operation, system function or response of plant equipment and provides the same level of protection as originally designed. The Drywell Floor Drain Sump system performs l               no active control functions nor is it capable ofinitiating any accident or equipment malfunction I
previously evaluated in the USAR. Therefore, this change does not create the possibility of an
              . accident or equipment malfunction of a different type. Technical Spcification Bases B3.4.7, "RCS Leakage Detection Instrumentation," indicates that the primary means of quantifying laakage in the Drywell is the Drywell Floor Drain Sump Monitoring system. This modification
                                                                     ~
l               will revise the Technical Specification Bases to add the pump alternating and Hi-Hi level alaim l               functions to the Programmable Logic Controller. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
CONDENSATE FILTRATION OPERATION AND PANEL REVISION Activity Evaluated:: ECN 27740 and USAR Change 8-183                            Log Number: 98-070
              - Engineering Change Notice (ECN) 27740 revises the backwash protocols with respect to the function of the prefiltering system, bypasses the defective flow meters in the sample panel with flexible hoses, and revises the building layout drawings. All of these changes for the condensate filtration system meet design, material, and construction standards. There at'e no credible failure
,i              modes related to this activity that can act as initiating events for any design basis accident analyzed in the Updated Safety Analysis Report (USAR). This activity does not change system performance or system response characteristics, and does not cause the system to operate outside its design limitsc The ' condensate filtration and condensate cleanup systenis are non safety related and are not required to function during any design basis accident or transient. Therefore, this
             ~change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR' Failure of the backwash mode would place the affected filter system into a bypass mode and send the condensate water directly 1into the appropriate condensate polisher. Failure of the flexible hoses would be contained within the sample panel. In the event that the amount ofleakage was not detected readily, the worst case scenario would be'that the water would fall onto'the floor and the floor drains in the
              . immediate area would capture the spillage. These actions are acceptable and could not create an accident or malfunction of equipment important to safety of a different type than previously
              . evaluated in the USAR.E This activity does not alter any acceptance values, safety limits, limiting
   'I i i

Attachment 3              '
to U-603276 Page 1I8 of 252 safety, system settings, or limiting conditions for operation for this system. Therefore, this activity doet. not reduce a margin ofsafety as defined in the basis for any Technical Specification.
DIVISION 1 NSPS DISTRIBUTION BUS IC71-P001 A LOAD RECORDING DURING PLANT OPERATION lA ctivity Evaluated: Temporary Modification 98-055.                              Log Number: 98-073
     . Temporary Modification (TM) 98-055 provides a means ofinstrumenting the Division I Nuclear System Protection System (NSPS) Bus IC71-P001 A loads during normal plant operation. This TM installs ring type current transformers (cts) at the NSPS Diodution Panel IC71-P001 A-
     ' load breakers. The equipment added does not perform any safety related function. The equipment is non-intrusive and has no active function which affects the ability ofNSPS bus loads to perform their intended function. During installation and removal of the temporary modification, equipment on the NSPS Bus, Divisic n 1 NSPS self test system, and Division 1 NSPS Bus IC71-P001 A is de-energized.' The inverter and plant safety systems are operated in a s normal manner per Clinton Power Station approved procedure. The installation and removal of the temporary modification is to be performed during plant Modes 4 or 5 only. The Division 1 NSPS inverter is powered from its normal DC power supply with the alternate AC source available. The monitoring of the loads is achieved through an isolated circuit which is non'-
intrusive to the NSPS bus loads. An evaluation of design criteria deviation concluded that the
     ; cts and associated circuits do not degrade the performance of the safety related circuits.
Therefore, this change does 'not increase the probability or consequences of an accident _or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis
     - Report (USAR). The addition of passive circuit elements does not ' disable the protective features of the NSPS bus feeder breakers. The evaluation of design criteria deviation also concluded that there are no credible' failure modes which can prevent the functioning of the NSPS distribution bus during normal and postulated accident conditions. The failure modes of the NSP.S power
     ' supply system are unchanged by the modification. If there is a loss of the NSPS bus, the resulting
                                                                                                                     )
affect is los's of Residual Heat Removal (RHR) shutdown cooling. USAR section 15.2.9                          ]
     - previously evaluated Loss of RHR shutdown cooling event.- Therefore, this temporary modification does not createi the possibility.of an accident or malfunction of equipment important
     ; to safety of a different type than previously evaluated. The Techni_ cal Specifications do not -
: describe details ofNSPS Distribution bus circuit configuration.~ Technical Specification 3.8.8 requires only one inverter in Modes 4 and 5. Division 2 inverter will be available to satisfy this Technical Specification. This change does not affect the i oltage and the output frequency of the
;. inverter. Therefore, this activity does not reduce a margi. . of safety as defined in the basis for any Technical Specification..
l 2
h

l .                                                                                  Attachment 2 to U-603276 Page i19 of 252 RESIZE SX/RHR HEAT EXCHANGER ORIFICE TO INCREASE FLOW TO 5800 GPM l     Activity Evaluated: ECN 30860 and USAR Change 8-143                            Log Number: 98-074 Engineering Change Notice (ECN) 30860 installs a 7.3" orifice plate in place of the existing 7.0" plate in the Shutdown Service Water (SX)/ Residual Heat Removal (RHR) heat exchanger discharge line. This activity results in Updated Safety Analysis Report (USAR) Table 6.2 reverting back to original design to indicate 5800 gallons per minute (gpm) in lieu of 5550 gpm and indicating the original heat transfer coeflicient. The SX system is an accident mitigation system and is not assumed to initiate any design basis accident. The design basis accident / modes that would most )ikely be impacted is the Loss Of Coolant Accident - Post Accident Suppression Pool Cooling Mode. Heat transfer rate can affect containment response. For a given heat transfer coeflicient, the higher the SX flow rate the higher the heat removal. The only potential affect on the consequences of an accident would be to increase SX flow to the RHR heat exchanger :o the point where flow to other station auxiliaries are decreased to unacceptable levels. SX Division 2 already uses a 7.3" orifice and design flows through other SX loads have been met. The SX Division I design was verified by flow balancing SX Division 1 per Clinton Power Station Procedure 2700.12 prior to release for operation. Therefore, this change does not increare the probability or consequences of an accident or malftmetion of equipment important to safety previously evaluated in the USAR. This activity does not result in an appreciable change to the overall SX system characteristics or functions. As such, the change to slightly increase the size of a SX flow orifice to the RHR heat exchanger does not create an accident or malfunction of equipment important to safety of a different type than previously evaluated. Technical              J Specification Bases B 3.7.1 indicates 5800 gpm SX flow is supplied to the RHR heat exchangers.      l This activity ensures compliance with this Bases requirement. Therefore, this activity does not     i reduce a margin of safety as defined in the basis for any Technical Specification.
SIMULTANEOUS VERIFICATION OF METERING Ak < "OTENTIAL FUSES Document Evaluated: USAR Change 8-148                                          Log ' Number: 98-078 This change revises the several Updated Safety Analysis Report (USAR) sections to reflect the method of second verification when verifying Bus Metering and Potential fuses at Clinton Power Station (CPS). Simultaneous verification is required since a second, time-dependent verification would undo a portion of the initial positioning and would not meet the intent of an independent verification. Removal of these fuses can be independently verified, but during fuse re-installation the physical configuration of the Bus Metering and Potential Fuses and the associated doors / drawers require the verifier to undo a portion of the positioners action. The new method    l ensures that during fuse re-installation the Bus Metering and Potential Fuses are not mispositioned by the verifier. Since the final position and location of the fuses does not change, W change does not increase the probability or consequences of an accident previously evaluated w         AR. The likchhood of the bus metering and potential fuses being mispositioned is reduced by having two people concurrently verify that these fuses are in the proper position.
Therefore, this change does not increase the probability or consequences of an equipment

l i
I Attachment 2           ,
to U-603276 Page 120 of 252 malfunction previously evaluated in the USAR. This change represents administrative control over how the bus metering and potential fuses are verified upon restoration to service. This            j change makes no changes to component alignment, method of operation, or configuration of                4 components, therefore, this change does not create an accident or equipment malfunction of a different type. The Technical Specifications do not address specific methods for verifying the positioning of components; therefore, this activity does not reduce a margin of safety as defmed in the basis for any Technical Specification.
EXPERIENCE ASSESSMENT DEPARTMENT RESPONSIBILITIES Document Evaluated: USAR Change 8-149                                          Log Number: 98-079 This change revises Chapter 13 of the Updated Safety Analysis Report (USAR) reflect a change in responsibility for the processing ofIndustry Operating Experience. This change also reflects the organizational responsibilities for ownership of human error reduction efforts. In both cases, the responsibility changes from the Nuclear Safety and Performance improvement Depanment to the Experience Assessment Depanment. This USAR change also permits the use of approved site tracking systems to account for Centralized Commitment Tracking (CCTs) and Clinton Power Station (CPS) Task Tracking. This revision does not compromise or impact the compliance with seismic, fire loading, separation, or environmental design considerations for any structures, systems or components (SSCs) at CPS. This change does not compromise the performance of any system, introduce any new SSC, nor modify any existing SSC. Therefore, this change does not increase or create the probability or consequences of an accident or equipment malfunction.
This change affects the CPS organization at the manager and director / supervisor level, the operating crews and support organizations are not impacted. In addition, this change does not introduce any hardware changes. Therefore, the organizational change does not create the possibility of an accident or equipment malftmetion of a different type. Technical Specifications 5.1,5.2, and 5.3 address organizational responsibilities, mannin'g, and qualifications; this change does not compromise those responsibilities or required qualifications. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
CPS ORGANIZATIONAL CHANGE Document Evaluated: USAR Change 8-156                                         Log Number: 98-080 This changes revises Updated Safety Analysis Repon (USAR) Chapter 13 to incorporate title and organizational changes. The revision changes the following position titles within Plant Maintenance: Supervisor-Electrical Maintenance to Electrical Maintenance Manager, Supervisor-Mechanical Maintenance to Mechanical Maintenance Manager, Supervisor-Controls and Instrumentation (C&l) Maintenance to C&l Maintenance Manager, Supervisor-Maintenance Planning to Maintenance Planning Manager and Supervisor-Direct Suppon to Direct Support Manager. Also, the Supervisor-Maintenance Services changes to Director-Contract Services and the reporting path changes to Manager-Work Management. The position description for Manager-Work Management changes to clarify responsibilities. The position title for Director-i l
l l
l l
l l

U Attachment 2 to U-603276 Page 121 of 252 le Work Coordination changes to Director'-Work Management; identifying outage and on-line
;.              ! programs is added to the description of the position responsibilities. The positions for L                 Supervisor-Work Management and Supervisor-Scheduling are deleted, these positions report to
;               ~t he Director-Work Management whose responsibilities reflect the scope of these supervisor's
: l.                responsibilities. USAR Figure 13.1-1 is revised to reflect the above position changes. Position
            ,     responsibilities remain the same, but have been clarified. . The changes do not introduce any        .
structure, system or component (SSC), do not change or alter the functional performance
                - characteristics ~of any SSC, nor do they change the ability of the organization to mitigate an
                ; accident. Therefore, this change does not increase the probability or consequences of an accident or equipment malfunction previously evaluated in the USAR.' Further, the changes will not create a new or different type of accident or malfunction of equipment to safety different from those -
l               , previously evaluated in the USAR. ' Technical Specifications 5.1,5.2'and 5.3 address organization responsibilities, manning and qualifications; this change does not compromise those responsibilities or required qualifications: Therefore, this change does not reduce a margin of
                . safety' as defined in the basis for any Technical Specification.
REPLACEMENT OF EMERGENCY RESERVE AUXILIARY TRANSFORMER g                : Activity Evaluated: Modification AP-040 Supplement 6,-                          Log Number: 98-081 ECN 30897, USAR Change 8-221 -
This activity replaces the fixed tap Emergency Reserve Auxiliary Transformer (ERAT) nominally rated 138kV/4kV,- 15 MVA with 4 fixed no load taps with a new nominally rated 138kV/4kV,18 MVA (with no fans operating) with 17 taps which can be changed while the transformer is in
                 . service and carrying load. The change was made to respond to variations in the voltage available on the 138 kV grid. The new transformer's larger capacity of 30 MVA (with the fans operating) is necessary to support operation of the ERAT Static Var Compensator. The new transformer has t'wo sets of cooling fans which provide air flow to allow the transformer to carry the additional load. The new transformer also has the capability of adjusting the transformer ratio to respond to variations in the voltage available on the 138 kV grid. The new transformer is designed to the same voltage rating and a larger arrent rating than the replaced transf orraer.
              ,     This change improves the performance of the Liass 1E Auxiliary power (AP) system becat se the larger transformer is less likely to be overloadeo and wi's run cooler with less long term degradation of the insulation. The transformer can operate up to 18 MVA ofload, without the
                 ~ fans operating, which is sufficient to carry all safety related buses. The reliance on the Reserve
                 - Auxiliary Transformer for Offsite Power during approximately two week outage of the ERAT
                  . complies fully with the Technical Specifications during Mode 4 and 5. In these Modes, only one source of offsite and one source of onsite power is required. ;The tap changer is enclosed in a separate tank to prevent the carbonization of the oil from tap changes under load from
                                         ~
L contaminating the main. oil tank. A failure of the new tap changer and new transformer is' no more L                 ' likely than a failu're of the existing transformer. The load tap changer (LTC) improves the Class
                                                ~
lE AP system performance for a wider range of 138 kV system voltages. The design of the tap changer requires a failure of two switches, a mispositioning of two switches, or a combination of
                   ' the two to' allow the tap changer to continue to operate after a tap change is complete.

m                                                                                                                 .
Attachment 2            , )
to U-603276               i Page 122 of 252
        ' Procedures require one qualified area operator to operate two switches ~ and'a second qualified area operator to simultaneously verify correct operation of those switches to change a tap.
Therefore, this activity'does not increase the probability or consequences of an accident 'or.
malfunction of equipment important to safety previously evaluated in the USAR. The credible failure modes that could occur are high voltage, low voltage and loss of power. These failure modes have already been evalur.ted and are consistent with the present failure modes. The installation of the new transformers significantly reduces the probability of these failure modes.
       . The design of the new transformer and tap changer provides protection against_ both equipment failures and against operator errors. Therefore, this activity does not create the possibility of an t
accident or malfunction.of equiprnent imponant to safety of a different type than previously evaluated in the USAR. The minimum voltage required to avoid a transfer from offsite power to the diesel generators is defmed by the degraded voltage allowable values in Technical                       i Specification Table 3.3.8.1-L _ By replacing the transformer with the ERAT LTC, the 4.16 kV bus voltage is maintained at a sufficient value during normal operation such that during an accident, the minimum 4.16 kV Class 1E bus volta'ge is adequate and does not result in an unnecessary transfer to the diesel generators. This change also preserves compliance with other Technical Specification requirements for electrical power distribution, availability, and redundancy.
       < Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical
       . Specification _
       - TEMPORARY INSTALLATION OF SCAFFOLDING ON THE CONTAINMENT POLAR CRANE
         ' Activity Evaluated:' MWR D60883 -                                                Log Number: 98-082
[This activity installs and secures scaffoiding to the bridge girders of the Containment Polar Crane in support ofMaintenance We rk Request (MWR) D60883. The scaffold shall provide access to the Containment liner plate above elevation 850', so that painters may repair degraded coatings.
The scaffolding shall be erected in accordance with the scaffolding manufacture's requirements, applicable plant procedures, site safety standards, and OSHA requirements. The scaffold shall be erected, used and disassembled while the plant is in Mode 4. The Containment Polar Crane will not be used to lift loads while the scaffolding is in place, being erected, or being disassembled.
The Fuel Handling Accident is described in Section 15,7.4 of the Updated Safety Analysis Repon (USAR). The Polar Crane is not used to lift fuel assemblies, there are no fuel bundles in the upper Containment fuel reactor, and the reactor head and drywell head are in place. Since these conditions are not going to change while the scaffolding is in place, there are no credible scenarios for this activity to initiate a Fuel Handling Accident. There is_a possibility that individual scaffold components or tools ~ could be dropped while scaffold or work activities are being performed creating missiles.: Clinton Power Station Procedure 1019.04, " Foreign Material Exclusion (FME)
       - Program," precautions shall be taken. An industrial cover and vinyl coated mesh shall be installed
: over the upper containment pools. In addition, the individual items that could be dropped are relatively light' weight.' The installation meets the applicable design, material, and construction standards. This activity does not affect the cranes overall performance or cause adverse

  .                                                                                   Attachment 2 to U-603276 Page 123 of 252 interaction with other systems. The Containment Polar Crane, classified as Safety Class 3, is the only component impacted by the scaffold installation. The Containment Polar Crane is not required to perform a safety related function when the reactor is in Mode 4 condition. This activity does not change, degrade, or prevent the system from performing actions described or assumed in the accident analysis for mitigating the effect of an accident or transient. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of        .
equipment important to cafety previously evaluated in the USAR. The scaffolding structure is designed in accordance with the US AR and NUREG-0800 requirements for non-category I structures that may interact with category I plant structures or equipment. The scaffolding has been tested for load resisting capability. There are no new mechanical or electrical components included in this temporary installation. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously              ;
evaluated in the USAR. There are no Technical Specifications that apply to temporary installation         ,
of scaffolding or the structures, systems, and components affected by this activity. There is no          l impact on Technical Specification safety limits, safety system settings, or limiting conditions for       j r
operation during the placement of scaffolding on the Containment Polar Crane. Therefore, this             l j     activity does not reduce.a margin of safety as defined in the basis for any Technical Specification.      )
i CORRECTION AND CLARIFICATION OF DESIGN POWER LEVEL                                                        l Document Evaluated: USAR Change 8-158                                           Log Number: 98-083 l'
This change corrects conflicting statements in the Updated Safety Analysis Report (USAR) associated with the power levels identified and utilized in the Chapter 1.5 transient and accident          ,
analyses. USAR Section 15.0.5. A states " safety analyses and evaluations were made for a                  i 104.2% power level of 3015 thermal megawatts." Contrary to this statement, Condition Report (CR) 1-98-03-016 indicates that several analyses in Chapter 15 were performed at power levels other than 104.2%. This change revises the wording in Section 15.0.5.A and other sections associated with the power levels to climinate this inconsistency.
The evaluation which resulted in this USAR change determined that all transients and accidents evaluated in Chapters 6 and 15 were performed correctly and in compliance with all requirements.
This change does not impact the failure modes or the initiating events. These changes ensure that the descriptions of the analyses as provided in the USAR are consistent with the analyses as performed. The plant design basis is correct and no systems operate outside of their design limits as a result of these changes. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. This change has no direct or indirect impact on structures, systems, or components. The equipment assumed to function in the transient and accident analyses continue to function in the same manner and with the same reliability. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This change does not introduce any new equipment or change any existing structures, systems, or components in the plant, This change corrects conflicting statements and provides clarification as to the basis for the initial power level used in the transient and accident analyses. Therefore, this change does

 'h                                                                                                                 .
Attachinent 2
* to U-6C 3276 Page 124 of 252 not create the possibility of an accident or equipment malfunction of a different tyoe. The design basis analyses remain valid and the changes do not impact those analyses. These changes are                     !
consistent with the analyses as performed and are consistent with the Techni;al Specifications for              {
the plant.; Therefore, this activity does not reduce a margin of safety as defined in the basis for .
any Technical Specification.
        ' INSTALLATION OF HANDSWITCH FOR THE CONDENSER VACUUM PIJMP -
DISCHARGE VALVE Activity EvaluatedL ECN 30817; USAR Change 8-159;                                Log Number: 98-084
                                 . and ORM Change 25-3 Engineering Change Notice (ECN) 30817 installs a two position spring-return switch in the condenser vacuum pump circuit to allow the Condenser Separator Tank Discharge Valve to be manually opened prior to starting the Condenser Vacuum PumpJThe discharge valve is
        . co'ntrolled via an air operated solenoid and will fail close on loss of air or loss of power. The Condenser Vacuum system is non-safety related and powered from a non-safety related power sourceJ The only probable time that a transient could be initiated would be when steam is being admitted to the condenser, a vacuum is present, and the vacuum pumps are being prepared to be staned.- The transient would be limited to a loss of condenser vacuum but a check valve is in place upstream of the pump;to prevent inleakage, and therefore, limit the actual vacuum loss. The area of concern in this entire change is the radiological consequences of manually holding the .
valve open or failure of the switch such that the automatic closure upon receipt of a main steam line high-high radiatiori signal is defeated. Per the design basis analysis, all of the airborne activity that was transported to the condenser is assumed to leak directly to the environment at a rate of LO% per day. Per General Elutric calculations, the condenser leak rate is assumed to be 1% per e      day with no credit being taken for isolating the mechanical vacuum pumps. By bypassing the
      ' s isolation of the discharge valve for a few minutes or longer, there is no increase in the condenser leak rate, and no increase in the radiological cons'equences to the environment. This activity does not affect the environmental, seismic or separation criteria of any system important to safety. No new testing requirements have been imposed by this plant change. Therefore, this activity does
        ~ not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (US AR). Failure of any component in the Condenser Vacuum system does not initiate any sequence of events resulting in any evaluated accident. Failure of the switch contacts in the open position would result in the discharge valve not opening.' Failure of the switch contacts in the closed position would result in the discharge valve opening and/or remaining open. Failure of the new .
,         handswitch has no affect on the ability of the plant systems to perform their safety function.
These malfunctions are bounded by existing equipment malfunctions. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a
        . different type than previously evaluated. There are no Technical Specifications associated with
        . the Condenser Vacuum system, including its logic and controls for the system. This activity does not degrade any Technical Specification safety limit or limiting condition of operation, and

i                                                                                                  Attachment 2 to U-603276 Page 125 of 252 l;.
L                    therefore, does not reduce a margin of safety as defmed in the basis for any Technical LSpecification.-
AR/PR MONITOR REDUCTION Activity Evaldated: Mdd PR-040, ECN 30790, -                                     Log Number: 98-085    -
                                               ' USAR Change 8-177 and Eplait Revision Plant Modification PR-040 replaces the central area and process radiation monitoring system in
                    ' the Main Control Room (MCR), Technical Support Center, and Radiation Protection Office.
                  ; Also, the modification reduces the number of monitors in order to reduce the possibility of
                 , electrical noise on the six communication lines. 31 monitors and 38 remote ports from the Central Control Terminal (CCT) in the MCR have been disconnected. A review of the Updated Safety Analysis Report (USAR) determined that the monitors being disconnected provide no information necessary for mitigating any of the design basis accidents. By reducing the number of monitors connected to the CCT, the performance of the CCT will improve as a result of electrical noise reduction.~ None of the monitors being disconnected are required by Regulatoiy Guide 1.97.
USAR Section 7.6.1.2.2.5.1 statei 'Each channel's safety-related functions are designed to operate without the use of any of the non-safety related ponions of the radiation monitoring            i system." Since the operation of the CCT is not being changed and the operation of the monitors in a stand-alone mode has not changed, there is no increase in the probability or consequences of-
                                                                                  ~
                  < an accident ci malfunction' of equipment important to safety previously evaluated in the USAR.             )
The significant failure modes are failing to alarm when exposure rates are exceeded or alarming inappropriately, Procedures exist for the periodic verification oflocal radiological conditions L independent of the monitors and for channel and source checks to ensure monitors continue to functionf A failure to alarm would be detected as a result of these periodic verifications.
L                      Therefore, this activity does not create an accident or malfunction of equipment imponant to l                       safety of a different type than previously evaluued. The information provided by the Area Radiation / Process Radiation (AR/PR) system and the CCTs is not used by the ' operators to determine the proximity to any Technical Specification safety limit. The monitors being disconnected are not addressed in the Technical Specifications. Therefore, this activity does not l:                      reduce a margin of safety as defined in the basis for any Technical Specification.
1
: SCREEN WASH DECHLORINATION SYSTEM
                     ' Document Evaluated! Modification CL-008; USAR Change 8-192                       Log Number: 98-086
                     ' Modification CL-008,provides a 'shdiuo bisulfite system to dechlorinate screen wash water effluent in the Circulating Water Screen House (CWSH). The dechlorination system is designed to' automatically feed diluted sodium bisulfite when any of the' traveling screens are being washed    j in automatic or' manual control. This dechlorination system is specifically intended to satisfy the
                      ' National Pollutant Discharge Elimination System (NPDES) permit requirements for Service L                     l Water discharges via this path (outfall 006). The failure of the screen wash dechlorination system cannot act as an initiating event for malfunction or failure of any stmeture, system, or component. l

          <     g Attachment 3
* to U-603276               l Page 126 of 252
     <               This modification meets design, material and constiuction standards applicable to installation of a
: non-safety related system in the CWSH. The only interface the dechlorination system has is to inject a chemical into the screen wash discharge before it leaves the plant. Therefore, this change
                  ' does not increase the probability or consequences of an accident previously evaluated in the
                   . Updated Safety Analysis Report (USAR). The injection point for the sodium bisulfite is Jdownstream of all plant equipment. The consequences of a spill or leak would be minor. Sodium bisulfite is neutralized by water, Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The failure of the screen wash dechlorination system could not impact any safety related system.1 The changes are limited to non-safety related systems and are located in non-safety related areas. Therefore, this change does not create the possibility of an accident or equipment malfunction of a different' type. The Technical Specifications do not discuss any system that interfaces with the screen wash dechlorination system.' Therefore, this activity does not reduce a margin of safety a's defined in the basis for any Technical Specification.
: INSTALLATION OF hTATIC VAR COMPENSATOR FOR RESERVE AUX 1LIARY TRANSFORMER.
                    ' Activity Evaluated: Modification AP-037, ECNs 30526,                        Log Number: 98-087 30527,30528 and USAR Change 8-205 Engineering Change Notices (ECNs) 30526,30527, and 30528 provide for installation of the Reserve Auxiliary Transformer (RAT) Static Var Compensator (SVC). The SVC is designed to o   >
                  ; quickly and efficiently change _the total impedance of the Offsite Power System to match the capacitance and inductance. The use of the SVC to match the capacitance and inductance
{
                  ' reduces the reactive component of the current to a negligible value and leaves only the real             q current component flowing through the RAT; thus, lowering the voltage drop and raising the
                  = voltage at the 4kV terminals of the RAT. The SVC can accomplish this task very quickly and accurately by use of reactors, capacitors, and thyristors. The equipment and control building for
                  ' the RAT SVC are being installed on foundations which were designed and installed under modification AP-0361 The power for the RAT SVC is ' supplied from two sources: one from a
                   . 4kV-480 volt transformer in the SVC yard and the other from a 480 volt Balance of Plant source which is normally fed from the Urdt Auxiliary Transformer or the RAT.
L The only work authorized by this modification is: the mounting and wiring for SVC switches and meters in the Main Control Room (MCR), the additional wiring for the auxiliary contacts from the safety related and non-safety switchgear, and the power supply for station power. The new design for the wiring and meters mounted ~on MCR panel P870 meets the same standards as the             ,
rest of the circuitry. The installation of these components could introduce some risk of human          j error causing a Lbss of Offsite Power (LOOP) since both RAT and Emergency Reserve Auxiliary             l Transformer (ERAT) controls are located on P870. No connections are being made from RAT to
                   . ERAT circuits an'd the new wiring uses existing terminal blocks and wiring.. The installation of the RAT SVC will be performed in accordance with plant procedures. The installation of the J additional wiring'for the auxiliary contacts from the safety related and non-safety switchgear will i
,                                                                                                                            l L
L    .,

p' 4
   .                                                                                  Attachm:nt 2 to U-603276 Page 127 of 252 be done both outside of an.d during divisional outages, but will have little or no interaction with a RAT or ERAT outage. The station power design brings power from the plant to the SVC. No connection will be made between the plant and the SVC until Nuclear Regulatory Commission approval has been received for requested Technical Specification changes. This activity does not affect any safety-related systems ability to perform its intended function , nor unnecessarily challenge the system during installation. The scope of work involving wiring from divisional components to the optical isolators will be perfonned within the scope of divisional outages with appropriate Technical Specification Limiting Conditions for Operation (LCOs) and with risk assessments as defined by plant procedures. Therefore, this activity does not increase the            j probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). Since the wiring, cabling, meters and switches meet existing design standards (in particular, for electrical separation), this activity does not introduce any new failure modes. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The installation of the RAT SVC has no affect on the DG nor Onsite Power (AP) systems' ability to perform their Technical Specification required function, and    l thus, have no affect on any acceptance values as long as no connections are made between the           !
plant and the RAT SVC. Therefore, this activity does not reduce a margin of safety as defined in      l the basis for any Technical Specification                                                             l l
CHANGE OF DESCRIPTION TO BUS DUCT LINE UP DURING ERAT OUTAGE                                           j I
Document Evaluated: US AR Change 8-161                                         Log Number: 98-089 This change revises the specific line up of the non-segregated bus duct switches used during an l      Emergency Reserve Auxiliary Transformer (ERAT) outage described in Updated Safety Analysis Report (USAR) Section 8.3.1.1.2. The current line up leaves the plant with no Balance of Plant power and subjects the plant to more risk than necessary. This change revises the disconnect switch positions as follows: RTl4 Open to RTl4 Closed and ET14 Closed to ET14 Open. This allows the non-safety related and safety related busses to be fed from the Reserve Auxiliary Transformer.(RAT) during an ERAT outage while maintaining power from the RAT to the non-safety busses.
This change assures a reliable source of power to safety and non-safety related busses and presents less overall plant risk since non-safety equipment is still operable. The only potential accident involved with this change is a loss of offsite power. The switches are lined up in a manner similar to their lineup during normal plant operations and leave the RAT as the source of l      power to the safety busses.- The only difTerence to this lineup is the safety busses are fed from the
!      RAT through the RAT bus which also feeds the non-safety busses. This change does not alter the j      potential causes of a loss of offsite power, nor does it alter the response of the unit to a loss of offsite power. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. All bus ducts involved are of the same design standards, so a trip of the RAT is not likely. Normally available protective devices will still be present and be capable of providing protective functions as designed. No hardware changes l
L

Attachment 2
     ~
to U-603276 Page 128 of 252 L beyond the switch positions are involved in this change. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously
   -     evaluated in the USAR. No new equipment is being introduced and the lineup is similar to the 1 lineup used when the ERAT is in service. Therefore, this change does not create the probability of an accident or equipment malfunction of a different type. This change does not affect any E
acceptance values for the AC power distribution or the Diesel Generator systems. This change             -
allows continued power supply to plant support systems consistent with normal operating conditions. ThereSre, this activity does not reduce a margin of safety as defined in the basis for any Technical Specmcation..
USAR' CHANGE TO CLARIFY STATEMENTS CONCERNING WASTE PR.OCESSING detivity Evaluated: USAR Change 8-173                                          Log Number: 98-090 These changes to the Updated Safety Analysis Report (USAR) clarify the current method of operation of the mobile solidification system. This change makes the follow'mg revisions to the USAR:' revises a statement in Section i1.4.2.2 to reflect that wastes are recirculated to the tank of origin not to the Spent Resin Tank; revises Section 11.4.2.3 to include the manifold between the system flanges and the flexible hose to connect to the mobile solidification vendor; and revises
        - Section'll 4.4 to replace cement and asphalt processed waste with dewatered wastes.-
The only accidents that the USAR discusses with respect to radioactive waste are the rupture of a radwaste collection tank and a drop and mpture of a liner of processed wastes. Neither of these accidents are affected by these changes.- The accident analyses for both of these accidents assume         ,
that the entire contents of the tank or container are released at once. If the manifold, hose or the recirculation line ruptured, the release would be spread over a period of time, and the cumulative volume of the release would be limited to the tank contents assumed in the analyses. Therefore, this change does not increase the probability'or consequences of an accident previously evaluated in the USAR. While the radwaste systems and the processing equipment are non-safety related
        . equipment, this equipment is important to safety for the prevention of release of radioactive material to the environment. The recirculation of the waste to the tank of origin uses the same            !
piping as the transfer system. The manifold is designed to the same specifications as the rest of the piping system. The containers used to rtore the dewatered waste have higher structure requirements than the dmms used for cement or asphalt processing. Therefore, this change does not increase the probability or consequences of an equipment malfunction previously evaluated in the USAR. These changes are administrative in nature and do not change the method of
       . processing any wastes. Therefore, this change does not create the possibility of an accident or equipment malfunction of a different type. There are no Technical Specifications associated with waste packaging or storage. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.

r-                                                                                                             1
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4
   .-                                                                                 Attachment 2              l to U-603276              ]
Page 129 of 252          l 1
1 j
TEMPORARY POWER TO TURBINE BUILDING RISERS 1 A AND 1C l
Activity Evaluated: Temporary Modification 98-062                             Log Number: 98-091 Due to electrical breaker refurbishment on the 6.9 kV bus l A (which feeds Unit Substation U),         f l
Temporary Modification (TM) 98-062 provides temporary power to Turbine Building Risers l A and 1C from 480 V Unit Substation 1K during the 6.9 kV bus l A outage. This temporary modification only changes the supply source of non-safety power for a non-safety load of the             l Auxiliary Power System. This temporary modification alters the electrical power configuration           ;
depicted on Updated Safety Analysis Report (USAR) Figure 8.3-1. Turbine Building Risers l A            1 and IC are not required to function in any but normal station operating conditions. The new temporary source of power performs in the same way as the normal source. The Turbine Building Risers operate the same from either source. All loadings and short-circuit contributions are within     ;
equipment and cable ratings. The components and equipment powered from the buses and risers            j are not credible initiators of any accident evaluated in USAR Chapter 15. Therefore, this change        i does not increase the probability or consequences of an accident or malfunction of equipment            j important to safety previously evaluated in the USAR. Breaker tap settings provide overload            {
protection as well as coordination with upstream breakers. Since electrical loading is within           )
ratings and the breakers are coordinated properly, all equipment failures would occur in the same manner as before with similar consequences. Therefore, this change does not create the possibility of an accident or malfunction of equipment of a different type than previously              j evaluated. The load that this temporary modification affects is non-safety related and does not        {
I impact any Technical Specification requirement. The equipment continues to perform its intended non-safety related function. Therefore, this activity does not reduce a margin of safety as defined in the basis fer any Technical Specification.
NSED DEPARTMENT REORGINIZATION Activity Evaluated: USAR Change 8-165                                         Log Number: 98-094        l l
This change makes organizational changes to the Nuclear Station Engineering Department (NSED) as described in Section 13.1.2.1.2 of the Updated Safety Analysis Report (USAR). The            j changes include: disbanding the Resource Management Group and integrating its functions into ine organizations, forming an Engineering Support Group, moving NSED Work Management to Project Engineering to provide ownership of engineering planning and scheduling by the line           ;
organization, and moving Drafling to the Design Engineering Group. No responsibilities have been added or removed from the scope of responsibilities for the department. The responsibilities for these activities have only been moved from one Director to another. The qualifications,
!         experience, education requirements, of the engineers, directors, and support personnel are not i
being reduced. Engineering department personnel are generally not directly involved in operation or manipulation of plant structures, systems, or components. This activity does not alter the methods of performing engineering suppoit functions to the plant. This activity does not physically alter any structure, system or component. Therefore, this change does not increase or create the probability or consequences of an accident or equipment malfunction. Technical

I Attachment 2 to U-603276 Page 130 of 252 Specifications 5.2 'and 5.3 address manning and qualifications, this change does not compromise
                                                                                          ~
those stamng requirements or required qual.ifications. Therefore, this' activity does not reduce a margin of safety as defined in the basis for any Tec!mical Specification.
CHANGE IN TIME REQUIREMENT FOR THE DISTRIBUTION OF NRAG MEETING
: MINUTES AND REPORTS OF REVIEWS                                                                            -
Activity Evaluated ( ORM Change 21-7           .
Log Number: 98-095 This activity revises the time requirement associated with distribution of Nuclear Review and Audit Group (NRAG) meeting minutes and repons ofreviews to the Vice President / Chief Nuclear Omcer is beinp snged from "within 14 days" to "promptly" following each meeting and following completion of the reviews. The NRAG function and responsibilities are not impacted by this change and their abilities to identify and implement prompt, effective corrective actions still remains. No system, structure, component, or function is affected or implemented by this change. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment impodant to safety or create one of a d_ifferent type than previously evaluated in the Updated Safety Analysis Repon. Technical Specification 5.2.I requires that onsite and offsite organizations be established for unit operation. This requirement is not being changed.' The function and responsibilities of the NRAG are not impacted by this change.
Therefore, this change does not reduce a margin of safety as defined in the basis for any Technical '
Specification.
          . EXPAND EDUCATIONAL AND EXPERIENCE Rl;QUIREMENTS FOR NSED
         ~ ActivityEvaluated: USAR Change 8-167                                          Log Number: 98-096             l Updated Safety Analysis Report (USAR) Change 8-167 expands the acceptable scope of the                       i educational and experience requirements for technical support personnel working in Nuclear                   i Station Engineering Department (NSED). The activity consists of revising USAR Section 13.1.1.1.3 by substituting the statement "NSED positions may be filled by either a degreed engineer or a non-degreed individual with equivalent experience in the area of assigned responsibility. Individuals in NSED without engineering degrees are classified as Specialists" for tb Matement " Entrance level engineering positions, for example, require a college level qineering degree." A possible failure mode, where the change in education / experience requirements might act as an initiating event,'would be if an NSED individual (without an engineering degrce) performed an engineering task that resulted in a poor design of plant equipment used to mitigate accidents or transients. 'NSED staff, who perform independent work                j on' systems, must be certified to' perform tasks in accordance with the Engineering Suppon (ES)
Training Program Description (TPD).. The existing' administrative barriers of training and
_ qualification remain in place.1 Activities or tasks contained in the ES TPD, identify the required prerequisite level of education'and experience. This administrative barrier is reinforced by                   .
         . additional technical and supervisory reviews of NSED products.. Plant equipment is not affected
         ~ by the proposed change and thus the change has no radiological impact. Therefore, this activity 4
j' 3                                                                                                               j

r
  .                                                                                   Attachment 2 to U-603276 Page 131 of 252 l
does not increase the probability or consequences of an accident or malfunctica of equipment important to safety previously evaluated in the US AR. This activity does not introduce any new structures, systems or components, failure modes or mechanisms for failure of equipment                 j important to safety. Therefore, this activity does not create the possibility of an accident or         I malfunction of equipment important to safety of a different type than previously evaluated. This        I USAR change does not involve any Technical Specification Limiting Conditions for Operation,           -
I acceptance values, or design limits. This change is consistent with the qualification requirements established in Section 5.3, Unit Staff Qualifications, given in the Improved Technical                   !
Specifications. Consequently, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.                                                                  ;
INSTALLATION OF STATIC VAR COMPENSATOR FOR EMERGENCY RESERVE AUXILIARY TRANSFORMER Activity Evaluated: Modification AP-038, and                                    Log Number: 93-097      l ECNs 30532, 30533, 30534                                                       I Engineering Change Notices (ECNs) 30532,30533, and 30534 provide for installation of the Emergency Reserve Auxiliary Transformer (ERAT) Static Var Compensator (SVC). The SVC is                 i designed to quickly and efficiently change the total impedance of the Offsite Power System to           I match the capacitance and inductance. The use of the SVC to match the capacitance and inductance reduces the reactive component of the current to a negligible value and leaves only the      I real current component flowing through the ERAT; thus, lowering the voltage drop and raising            I the voltage at the 4kV terminals of the ERAT. The SVC can accomplish this task very quickly             I and accurately by use of reactors, capacitors, and thyristors. The equipment and control building for the ERAT SVC are being installed on foundations which were designed and installed under modification AP-036. The power for the ERAT SVC is supplied from two sources: one from a                 l 4kV-480 volt transformer in the SVC yard and the other from a 480 volt Balance of Plant source           l which is normally fed from the Unit Auxiliary Transfonner or the RAT.
The only work authorized by this modification is: the mounting and wiring for SVC switches and meters in the Main Control Room (MCR), the additional wiring for the auxiliary contacts from the safety related and non-safety related switchgear, and the installation of the SVC Interface Panel.
The new design for the wiring and meters mounted on MCR panel P870 meets the same standards as the rest of the circuitry. The installation of these components could introduce some risk of human error causing a Loss of OfTsite Power (LOOP) since both Reserve Auxiliary Transformer (RAT) and ERAT controls are located on P870. No connections are being made from RAT to ERAT circuits and the new wiring uses existing terminal blocks and wiring. The installation of the ERAT SVC will be performed in accordance with plant procedures. The installation of the
;      additional wiring for the auxiliary contacts from the safety related and non-safety switchgear will be done both outside of and during divisional outages, but will have little or no interaction with a RAT or ERAT outage. The station power design brings power from the plant to the SVC. No connection will be made between the plant and the SVC until Nuclear Regulatory Commission approval has been received for requested Technical Specification changes. This activity does not I

                                                                                            ~
Attachment 2            I to U-603276              l Page 132 of 252 affect any safety-related systems ability to perform their intended function , nor unnecessarily challenge the system during installation. The scope of work involving wiring from divisional components to the optical isolators will be performed within the scope of divisional outages with appropriate Technical Specification Limiting Conditions for Operation (LCOs) and with risk assessments as dermed by plant procedures. Therefore, this activity does not increase the               i probability or consequences of an accident or malfunction of equipment important to safety              I previously evaluated in the Updated Safety Analysis Report (USAR). Since the wiring, cabling, meters and switches meet existing design standards (in particular, for electrical separation), this activity does not introduce any new failure modes. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The installation of the ERAT SVC has no affect on the DG or Onsite Power (AP) systems' ability to perform their Technical Specification required fimction, and thus, have no affect on any acceptance values as long as no connections are made between the plant and the ERAT SVC. Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
ELECTRONIC DOSIMETER SELF PERFORMANCE CHECKS Activity Evaluated: US AR Change 8-168                                          Log Number: 98-098 Regulatog Guide (RG) 8.28 states "Each day, before use, audible alarm dosimeters should be tested for adequate battery strength and should be checked with a radiation source for proper operation." An exception to RG 8.28 is the Merlin Gerin Model DMC-100 Electronic Alarming Dosimeter. The Merlin Gerin self performance checks ensure the detector responds to radiation counts, microprocessor and detector circuitry is functioning correctly, and batten voltage is acceptable. If these tests fail the applicable acceptance criteria, an error message is displayed.
The self checks performed by the Merlin Gerin DMC-100 are adequate in determining if the electronic dosimeter is properly operating without the use of a response check to a radiation           j source. This change revises Section 1.8 and Section 12.5.2 of the Updated Safety Analysis                !
Report (USAR) by endorsing RG 8.28 with exceptions.
Electronic dosimeters are used in monitoring personnel exposure, they are not equipment important to safety. This change does not involve operation, maintenance, or design of equipment        l nor does it affect the environment of that equipment. In addition, it does not impact any manual         l or automatic actions that would be taken in the event of an accident. Failure of the electronic          l dosimeter would not degrade or prevent equipment from performing actions described or                    I assumed in the accident analysis, nor would it affect any system important to safety such as to unpede access to vital areas. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously l
evaluated in the USAR. Failure of the electronic dosimeter does not introduce new failure modes          l or mechanisms for equipment important to safety. Therefore, this change does not create an accident or equipment malfunction of a ditTerent type. Electronic dosimeter performance checks are not discus:ed in the Technical Specifications. Technical Specification Section 5.7 requirements for controls over high radiation areas are not impacted by this activity and do not

Attachment 2 to U-603276 f      ,
Page 133 of 252
         . establish margins of safety as programmatically defined. Therefore, this activity does not reduce a L margin of safety as defined in the basis for any Technical Specifications.
PLANT RADIATION PROTECTION AND CHEMISTRY REORGANIZATION l
Activity Evaluated: L USAR Change 8-169                                          Log Number: 98-099     -
This Updated Safety Analysis Report (USAR) change is an administrative change depicting the j        ; reorginization of the Plant Radiation and Chemistry Group. This change involves the following
          ' changes: eliminates the Supervisor - Health Physics Analysis Group (HPAG) position and
;          incorporates the responsibilities of HPAG under the Supervisor - Radiologicd Support; and R           eliminates the Assistant Director Plant Radiation Protection position in its entirety. These changes alter the organizational discussion and hierarchy depictions given in USAR Sections 12.5 and 13.1. This organizational change preserves the existing HPAG functions and qualifications under the Radiological Support Group. It does not involve operation, maintenance, or design of equipment important to safety nor does it affect the environment of that equipment. Also, it does l
not impact actions that would be taken in the event of an accident or equipment malfunction.
l           Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. Nor does this i           change create a possible accident or malfunction of equipment important to safety of a different
                                                                                                     ~
type than previously evaluated in the USAR. Technical Specifications 5.1, 5.2, and 5.3 address organization, manning, and qualification requirements, this change does not compromise those l           responsibilities or required qualifications. Therefore, this activity does not reduce a margin of        i I
          - safety as defined in the basis for any Technical Specification.
ADD RESTRICTION ORIFICE TO SHUTDOWN SERVICE WATER (SX) RHR HEAT EXCHANGER BYPASS l
Activity Evaluated: ECN 30903; USAR Change 8-171                                Log Number: 98-101 L
!         ~ During operation of the Shutdown Service Water (SX) system, high flows through the SX/ Residual Heat Removal (RHR) bypass piping were observed. As a result, flow to some safety related coolers supplied by SX could be diverted to the SX/RHR bypass. This change installs a l
restriction orifice in the bypass piping flow path to decrease flow. This orifice is designed to pass approximately 3600 gallons per minute (gpm)._ This is above the required SX pump minimum                 i flow requirement of 3200 gpm continuous, but below the flow rate that could decrease flow to other coolers. These flow paths are depicted in the Updated Safety Analysis Report (USAR)
Figures 3.6-1, Sheet 47, and 9.2-2, Sheet 1. The SX consists of three divisions which correspond
            . to the three electrical safety divisions. Any two of these divisions operating together are adequate to assure safe shutdown of the station. This change, which adds a flow orifice to ensure the
            ~ USAR design flows 'are met throughout the SX system,'does not increase the probability or consequences of an accident previously evaluated in the USAR. This change reduces flow through the SX/RHR bypass to acceptable levels. Addition of the orifice assures that flows
           . 'through all heut exchangers meet design requirements.- The potential for too much hydnulic

e,
                                                 ,                                           Attachment 2 to U-603276 Page 134 of 252 resistance has been addressed by Engineering Change Notice calculations and verified by SX ll          ; system flow balance prior to release for operation. The revised piping and suppon design is in h
accordance with existing codes and standards. Therefore, this change does not increase the probability or consequences'of a malfunction of equipment important to safety previously l
           ' evaluat'ed in the USARJ Chapter 6 and 15 do not identify any accidents that SX is assumed to L initiate. The orifice being'added is a passive device. The additional piping and support design is        -
           . in accordance with existing codes and standards; as such, any passive failures are enveloped by existing analysis in USAR Table 9.2-5. No changes are made which impact the Moderate Energy Line Break analyses previously performed. Therefore, this change does not create an accident or equipment malfunction of a different type. The orifice modifies the system so that the design basis c is maintained in accordance with the USAR, the change does not result in a significant change to the overall SX system characteristics. This change contributes to the assurance that assumptions
           . regarding SX system performance le .:Is assumed in Technical Specifications 3.7.1 and _3.7.2 are fully satisfiedi Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
fADD RESTRICTION ORIFICE TO SHUTDOWN SERVICE WATER (SX) RHR HEAT EXCHANGER BYPASS Activity Evaluatedi ECN 30904; USAR Change 8-172                               Log Number: 98-102 During operation of the Shutdown' Service Water (SX) system, high flows through the                         ;
SX/ Residual Heat Removal (RHR) bypass piping were observed. As a result, flow to some safety
           ? related coolers supplied by SX could be diverted to the SX/RHR bypass. This change installs a i
restriction orifice in the bypass piping flow path to decrease flow. This orifice is designed to pass approximately 3600 gallons per minute (gpm). This is above the required SX pump minimum                     ;
flow requirement of 3200 gpm continuous, but below the flow rate that could decrease flow to                !
           - other coolersf These' flow paths are depicted in the Updated Safety Analysis Repon (USAR)
Figures 3.6-1, Sheet 47 and 9.2-2, Sheet 1. The SX consists of three divisions which correspond to the three electrical safety divisions. Any two of these divisions operating together are adequate to assure safe shutdown of the station. This change, which adds a flow orifice to ensure the                i USAR design flows are met throughout the SX system, does not increase the probability or                    l consequences of an accident previously evaluated in the USAR. This change reduces flow
: through the SX/RHR bypass to acceptable levels. Addition of the orifice assures that flows through all heat exchangers meet design requirements. The potential for too much hydraulic
: resistance has been addressed by Engineering Change Notice calculations and verified by SX                  !
I
           ' system flow balance prior to release for operation. The revised piping and suppon design is in accordance with existing codes and standards. Therefore, this change does not increase the probability or consequences'of a malfunction of equipment imponant to safety previously
           - evaluated in the USAR. Chapter 6 and 15 do not identify any accidents that SX is assumed to initiate.; The orifice being added is a passive device. The additional piping and support design is
           ' in accordance with existing codes and standards; as such, any passive failures are enveloped by
           ' existing ~ analysis in USAR Table 9.2-5. No changes are made which impact the Moderate Energy
: Line Break analyses previously performed. Therefore, this change does not create an accident or

Eo
' ~ .                                                                                      Attachment 2 to U-603276 Page 135 of 252 -
equipment ' malfunction of a different type. The orifice modifies the systr 3 so that the design basis is maintained in accordance with the USAR, the chnge does not result in a significant change to the overall SX system characteristics.' This change contributes to the assurance that assumptions regarding SX system performance levels assumed ic Technical Specifications 3.7.1 and 3.7.2 are fully satisfied. Therefo' re, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.        ,
: UTILIZATION OF WIDE RANGE AND FUEL ZONE LEVEL INDICATION DURING ~
EMERGENCY OPERATING PROCEDURE AND SEVERE ACCIDENT GUIDELINE USE Activity Evaluated: Emergency Operating Procedures /. -                          Log Number:' 98-103 -
_ Severe Accident Guidelines / Level Instrumentation Usage / CPS's Technical and Operational Position.
1 This activity discusses the position that the Fuel Zone Reactor Pressure Vessel (RPV) level instrumentation wi!! not be used for Emergency Operating Procedures / Severe Accident Guidelines
         -(EOPdAG) RPV level directed actions whenever Wide Range level instrumentation is still available. Essentially, Fuel Zone level instrumentation will not be used for RPV level actions until the RPV is depressurized. Clinton Power Station (CPS) does not have installed instrumentation to implement. Boiling Water Reactor Owner's Group Emergency Procedure Guidelines / SAG strategies when RPV level is below Top of Active Fuel (TAF) with the RPV still pressurized.
There is no requirement or expectation to modify the plant in order to provide such instrument capability. The entire range of Wide Range level instrumentation (+60' to -160") will be used.
The bottom of the Wide Range level instruments (-160") will effectively represent TAF (-_162',').
The Design Basis Accident - Loss of Coolant Accident (DBA-LOCA) serves as the envelope accident; all other accidents have less severe and less limiting level tracking requirements.
Utilization of Wide Range and Fuel Zone instruments as described in this change does not alter -
        ' the design response of the Emergency Core Cooling Systems (ECCS) and does not resuh in a change to the post DBA-LOCA reactor vessel level. The reactor vessel level instrumentatior. is not being modified by this change.; This position matches the design, material and construction standards for the reactor vessel level instrumentation. Utilizing the entire Wide Range level-instrument range, and only using the Fuel Zone level indication after Wide Range is no longer available ensure that RPV level is maintained above TAF, which is well above the DBA-LOCA event of 2/3 core height. Maintaining RPV level above TAF ensures that adequate core cooling is maintained, and therefore, the fission product barriers remain intact. The use of the entire Wide
        - Range water level instrumentation capability is a passive monitoring function only, and does not
        ~alter the equipment's safety related functions. Therefore, this activity does not increm,e the
_ probability or consequences of s.2 accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). Fuel Zone water level instrumentation is not safety related. These changes help minimize possible challenges to the ECCS/ Automatic Depressurization systems by minimizing the possibility of an operator having to initiate a premature RPV depressurization when RPV level is still being maintained above TAF.
L Therefore, this activity does not create the possibility of an accident or malfunction of equipment

e Attachment 2 to U-603276                f p                                                                                       Page 136 of 252
: important to safety of a different type than previously evaluated in the USAR. This activity is not a change to the Technical Specification safety limits, limiting safety system settings, or limiting conditions for operation; therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.                                                                 ;
f REPLACEMENT OF OBSOLETE LOGITROL PROGRAMMABLE LOGIC CONTROLLER                                           -
                                                                                                                   ]
       ' Activity Evaluated: Modification RT-038; USAR Change 8-191.                    Log Number: 98-104 Plant Modification RT-038 replaces the Reactor Water Cleanup (RT) Filter Demineralizer (F/Ds)'
Programmable Logic Controllers (PLCs) with new PLCs and an Operator interface touchscreen.
In addition, this modification also modifies Panel 1G36P002 to accommodate the new equipment and allows the semi-automatic operation of the panel switches to be performed by the Operator.
Engineering Change Notice (ECN) 27444 abandoned the dust collector in its place, otherwise, the
      - PLC logic sequence of events rernains the same. Although the Updated Safety Analysis Report (USAR) described functions in Section 5.4.8.1.2 are unaffected by this activity, USAR
      . Figure 5.4-18 is altered to reflect the new PLC configuration.
USAR Chapters 6 and 15 do not identify any accidents applicable to or potentially caused by the RT F/D system. This modification adds an operator interface and new PLCs to an existing panel on the 803' 3" elevation of the Containment Building..' However, the new equipment is above the
      ; pool swell elevation discussed ir, the pool swell accident analysis described in Chapter 6. The small quantity of hydrogen generating material introduced into the environment does not impact the assumptions and maximum quantities defined in USAR Table 6.2-49. Therefore, this change
: does not increase the probability of an accident previously evaluated in ' a USAR. This change affects a portion of a system providing no accident mitigatory function and does not otherwise arter system function.' Therefore, the consequences of previously evaluated accidents are not affected by this modification. This modification only affects the F/D portion of the RT system.
      )The F/D subsystem performs the cleanup portion of the RT system and has no safety related
      - function. The protective features, performance, and reliability of the RT system remain unchanged by this modification. Therefore, this change does not. increase the probability or               {
consequences of a malfunction of equipment important'to safety previously evaluated in the                 j USAR. The new PLCs do not introduce new failure modes or functions beyond those already                    j present; therefore, it d_oes not create the possibility of an accident or equipment malfunction of a different type. This modification introduces the touchscreen as an operator interface for operator convenience. If the touchscreen fails, the operator can still manually open and close the valves using the panel push buttons. Any mechanical damage to the screen resulting in breaking of the screen will not affect the suppression pool because the broken pieces will fall on the concrete floor only. Therefore, this change does not create the possibility of an accident or a malfunction         ,
ofequipment important to safety of a different type than previously evaluated. There are no                )
Technical Specifications associated with the RT F/D subsystem. The mention of the RT system in             l the Technical Specification Bases are for the use of the system for an alternate decay heat removal system. Also; the RT system is mentioned in reference to the Leak Detection instrumentation and detection functions.. The F/D subsystem does not perform this function. Therefore, this i
i L

i j
Attachmerit 2           !
to U-603276 Page 137 of 252 modification does not reduce a margin of safety as defined in the basis for any Technical Specification.                                                                                            .
REPLACE DIVISION I AND DIVISION 11 DG AIR START SOLENOID PILOT VALVES                                     ,
Activity Evaluated: ECNs 30744 and 30745; USAR Change 8-201                      Log Nt.mber: 98-105 l
Engineering Change Notice (ECN) 30744 encompasses Division I and ECN 30745 encompasses Division II. These plant changes ir lude the following: replaces existing Diesel Generator (DG)
Air Start Solenoid Operated Pilot Valves (SOVs) manufactured by Graham-White (G-W) with a new model manufactured by Automatic Velve (AV), removes DG Air Start Check Valves, removes and replaces the existing 3/8" diameter copper tubing with 1/4" stainless steel tubing           i using compression fittings, and revises the Piping and Instrumentation Diagram (P&ID) to show the existing 2-way check valves. The G-W SOVs are being replaced in response to a 10CFR21 notification concerning potential deficiencies in SOV operation under a4 ooerating conditions.
The other changes are enhancements or conforming changes to support SOV replacement.                     !
Chapters 6 and 15 of the Updated Safety Analysis Report (USAR) does not describe any design basis accidents isiated by failure of a DG. However, the DGs are designed to automatically start and be available to take load withir; 12 seconds of a start signal to support safety systems by supplying backup power in the event of a loss of off-site power. As such, the DGs are considered accident mitigation support. Failure of both the Division I and 11 DGs during a loss of off-site power constitutes a station blackout and could increase the consequences of an accident                  i previously evaluated in the USAR. However, it is established in the discussion below that the changes to the DG air start systems by these ECNs will not increase the probability of the DGs malfunctioning; and thus, do not increase the consequences of an accident previously evaluated in the USAR.
The new SOV has a smaller inlet and exhaust orifices than the existing valve. Since the flow rate is proportional to the cross-sectional area of the orifice, the decrease in orifice size by itself would be expected to increase the time for the starting air pilot pressure to reach operating pressures.
To offset this, the volume of the pilot air system is reduced by reducing the tubing diameter from 3/8" to 1/4" Calculation IP-M-0473 shows that the increase in time to reach AOV operating pressure is about 6 milliseconds. The calculation is based on the Technical Specification limit for air start system operability with a receiver pressure of 200 psig. This small increase negligibly affects the ability of the DGs to reach rated speed, voltage, and frequency within 12 seconds as required by the USAR and Technical Specifications. The AV SOVs are rated for 16 Watts versus 13 Watts for the existing G-W SOVs. The additional 3 Watts per valve results in a totalincrease of 12 Watts per Division and revises the load on the batteries. The new amperage remains well within the batteries' rated capacities and negligibly afTects battery loading. The Air Start Check Valves serve no function in this system and are bypassed in the present design. The function of the 2-way check valves has not changed and the valves are original equipment that have always been installed on the DGs. Therefore, these changes do not increase the probability or l
I l

1 l
Attachment 2 to U-603276 Page 138 of 252 L      consequences of a malfunction of equipment imponant to safety previously evaluated in the
               .USAR.
The changes to the diesel generator air start system maintain the capability ofthe DG to start                 )
               . reliably and power the safety busses. These changes do not affect any of the primary barriers to               i radiological release or any direct aceident mitigating systems. The new SOVs function the same               t l
              ; as the SOVs being replaced. ' Replacing the copper tubing with stainless steel tubing meets all i
fdesign and _ material requirements for control power piping and will improve the burst strength of the piping. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety of a different type The Technical' Specifications pertinent to this change include 3.8.L7 (the 12 sscond stan) and the
        >       . bases for 3.8.3.4 (multiple starts fiom air receiver pressure of 200 psig). The system changes preserve system operability in suppon of these Technical Specification requirements. The reliability of the Division I and II DGs to stan and provide timely backup electrical power is 'either preserved or enhanced by these changes. Therefore, these changes do not reduce a margin of safety as defined in the basis for any Technical Specification.
STORM DRAIN ADJACENT TO THE RADWASTE BUILDING -
               ! Activity Evaluated: CR l-97-02-203 CIEL Item 90-3                               Log Number: 98-107 Configured item Exception List (CIEL) Item.90-03 allows installation of a drain, connection to the existing storm drain, and minor grading in front of the east rolling steel door north of the -
    .            Radwaste Building. Condition Repon (CR) 1-97-02-203 identified that these changes had not
               . received an appropriate screening and, if necessary, an evaluation as required by ;10CFR50.59.
Updated Safety Analysis Report (USAR) Section 9.3.3.2.1 discusses the drainage of non-radioactive wastes is arranged within structural barriers to preclude contamination from neighboring radioactive areas.. USAR Section 2.4.12 and 2.4.13.3 discuss and evaluate the interaction of radwaste effluents and ground water. The Technical Specifications discuss limiting a      : the amount of radioactivity in outdoor tanks that are not surrounded by a dike. Thus, putting a          -
storm drain adjacent te the power block area should be evaluated.
USAR S ction 15.7.3 describes an accident thet involves the release of radioactive liquid in the iRadwaste Building. This accident actually involves the release of radioactive gas from spilled
                . liquid. The storm drains can have no effect on either the probability or consequences of this type of releasei Section 15AA5.3 describes small spilis and leaks of radioactive materials in which the offsite' dose will be negligible in comparison to the dose resuhing from the accountable plant leakage. ;Also there are floor drains in the Radwaste Buildingjust inside the rolling steel door
                .where the stonn drain was i_nstalled which will effectively prevent the release of the radioactive L liquid to the ' storm drain. USAR Sections 2.4.12 and 2.4.13.3 state that there is no possibility that effluems can move out the Radwaste Building duc to high ground water elevation. Further, this change does not create a potential unmonitored release pathway because ofits exterior location.
Therefore, this activity docs not increase the probability or consequences of an accident or

4 Attachment 2 to U-603276 Page 139 of 252
                          .       .                  a_
c malfunction of equipment important to safety previously evaluated in the USAR. This is a configuration change.which is limited to the storm drain system and its effect on the accumulation
               ' of rain water on the pcvement outside the Radwaste Building. There is no credible way an v               . additional storm drain at this location can initiate an accident involving the potential release of radioactivity. Therefore, this activity does not create the possibility of an accident or malfunction
               - of equipment important to safety of a differer.1 type than previously evaluated. The storm drains        .
                . are not discussed in the Technical Specifications. The Radwaste Building and its drainage
                . systems are already designed to preclude the release of effluent to the ground water system, and
               . this change is designed to preclude the entry of storm water into the Radwaste Building.
                . Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification?
REPLACEMENT OF MSR START UP DRAIN PIPS AND VALVES Activity Evaluated: ECN 30833; USAR Change 8-233                                Log Number: 98-108
                 . Engir.eering Change Notice (ECN) 30833 makes the follow'mg changes to Moisture Separator
                 - Reheater (MSR) start up' drain system: replaces the carbon steel material of the 1.5" diameter shell start up drain with chrome-moly material; replaces the carbon steel material of the bypass lines around shell . start up' drain valves with chrome-moly material; replaces all four shell stan up
: drain valves with body and bonnet material of chrome-moly material; and updates the vendor manual l mechanical department line list, mechanical department valve list, and drawing M05-1011-2.: The carbon steel material is susceptible to flow accelerated corrosion (FAC) while draining saturated ' condensate from MSR during start up. Chrome-moly material has a much higher ability'to resist damage from FAC. The turbine drain system removes condensate from the turbine steam path during start up and shut down when significant condensate is generated. This
                  '' drain path prot.ects the turbine from water induced damage. This system does not interface with any safety related system, nor does it influence an anticipated operational transient or postulated
                                                              ~
design basis accident previously evaluated. Failure of the valve to operate will eventually increase moisture content of the steam during start up or shut down. Also, rupture of the pipe would cause steam leakage inside the turbine building. Neither of these failures is considered to have an 7"
: increased probability, because no design or operational parameters are altered and system
                 . characteristic and performance requirements are not degraded. When the MSRs generate -                   J saturated condensate which flashes.into steam in the drain line, corrosion and erosion of the pipes    !
l
                 ' and valves occur. Chrome-moly material improves reliability and prolongs the life of the drain f system. - Therefore, this change does not increase the probability or consequences of an accident
,                    or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). Since no design, operational, or performance requirements (other than
: upgrading the material) are changed, the only credible failure modes due to this change would be        j failure of the valve to operate or a complete rupture of the pipe during plant operation and loss of isolation capability of the steam. However, these failure modes are the same as the original design failure modes. Therefore, this change does not create the possibility of an accident or malfunction    l of equipment imponant to safety of a different type than previously evaluated in the USAR. The l
Technical Specifications do not discuss the MSR startup drain system, nor is the system credited       j l
4 1
c I

i Attachment 2                   I to U-603276 Page 140 of 252 with providing a margin of safety supporting any other Technical Specification requirement.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
SHUTDOWN SERVICE WATER (SX) VACUUM BREAKER INSTALLATION Activhy Evaluated: ECN 30855, USAR Change 8-182                                Log Number: 98-109 Engineering Change Notice (ECN) 30855 modifies the Shutdown Service Water (SX) system to include twc vacuum breakers in series on the inlet side of the Division 111 Switchgear Room cEller condenser (IVXO6CC) and two vacuum breakers in series on the outlet side of IVXO6CC downstrearn of the chiller outlet isolation valve (ISX-025C). The outlet of the vacuum breakers are routed to a floor drain. Each line has one normally open isolation valve to facilitate routine maintenance on the vacuum breakers. Also, this change reroutes the discharge line for relief valve ISX-154C to tie into the SX piping downstream of 1SX-025C.
Chapter 6 or 15 of the Updated Safety Analysis Report (USAR) does not discuss the SX system as an initiating event for any accidents. This change does not alter the evaluated failure modes or          j modes assumed in the USAR accident analyses. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The SX system safety function is to transfer heat loads to the Ultimate Heat Sink under normal operating and accident conditions. On a three division SX system basis, the single failure of any one vacuum breaker to open potentially causes the failure of only one SX division. The SX system is designed such that failure of one division does not affect the other two divisions. Since the series / parallel        .
arrangement of the vacuum breakers on the Division Ill SX does preclude water spray / loss, since the design meets the single failure analysis assumptions of USAR Tables 9.2-5 and 9.2-6, and -               ;
since these components meet the material, design, and installation requirements for this system;             j this change does not increase the probability of a malfunction of equipment important to safety previously evaluated in the USAR. This change does not alter the assumed failure modes of the SX system beyond those already analyzed in the USAR. All equipment important to safety will                 )
continue to function as designed in support of plant operation and safe shutdown, therefore, this change does not increase the consequences of a malfimetion of equipment important to safety.
The installation of new valves and piping will be accomplished in accordance with approved specifications, codes and procedures. These components, therefore, are within the bounds of the passive failure analysis that was performed on the SX system in Table 9.2-5. Therefore, this change does not introduce a new or different type of accident or malfunction than any previously             )
evaluated in the USAR. Technical Specification (TS) 3.7 discusses the Division III SX system.                i Surveillance Requirement (SR) TS 3.7.2.1 does not apply to valves that cannot be inadvertently               I I
misaligned, which applies to vacuum breakers and relief valves installed by this ECN. SR TS 3.7.2.2 does not apply to this ECN because it does not affect any of the initiation logic cr components that are required to function in the event of an actual or simulated initiation signal.
By design, this change conforms to other requirements needed to support operability of the SX system. Therefore, this activity does not reduce a magin of safety as defined in the basis for any Technical Specification.

Attachment 2 to U-603276 Page 14I of 252 REPLACE ACTUATOR ON RHR PUMP MINIMUM FLOW VALVE TO INCREASE TORQUE THRUST Activity Evaluated: ECN 30932                                                   Log Number: 98-112 Engineering Change Notice (ECN) 30932 replaces the motor operator for motor-operated valve                .
      ' (MOV) lE12-F064A, Residual Heat Removal (RHR) Pump 1 A Minimum Flow Valve, with a larger motor to increase actuator torque / thrust capability. The original assumptions for actuator efliciency used for actuator analyses, made by the valve operator manufacturer Limitorque, were non-conservative. Limitorque Technical Update 98-01 and its supplement, reduced allowable actuator efficiency based on industry data. To compensate for the lower allowable efliciency of the actuator, a larger motor is being installed to provide 10 foot pounds of torque as opposed to the original 5 foot pounds.
Chapters 5,6, and 15 of the Updated Safety Analysis Report (USAR) discuss the Residual Heat Removal (RHR) system for mitigation of any accident that ultimately results in a plant trip. This new motor can not initiate any of the postulated accidents that might lead to the plant trip condition. Consistent with the current actuator, the larger motor will not over torque the valve           {
internals when the torque switch settings are within the requirements. Also, the larger motor has no adverse affect upon the Motor Control Center (MCC) bus loading or diesel generato loading               !
and is seismically acceptable. The motor change has no adverse affect upon the minimum flow valve or RHR system reliability, function, or method of performing its function. The minimum flow valve failure represents a single failure that has already been accounted for and, as a result, the dose consequences are not adversely affected. ' Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the US AR. When the flow in the main flow path falls to 750 gallons per minute, the minimum flow valve for the RHR A pump automatically opens and establishes a flow path to the suppression pool to prevent overheating and possible damage to the pump. The new motor has been procured to the same safety class as the original motor. This plant change does not interface with any other plant systems, and does not affect the design basis, function, or method of performing the functions of the RHR system.
Additionally, this plant change does not affect the design basis, function or reliability of the valve.
!           It is a configuration change to compensate for the lower allowable efliciency of the actuator. If
       - the valve fails open, the amount of coolant injected would be below the amount required in the USAR. If the valve fails closed during pump operation, it could damage the pump. Installation of the larger motor helps to ensure the valve performs its design basis safety-related function.
Therefore, this change does not increase the probability or consequences of a malfunction of              I equipment important to safety previously evaluated in the USAR. The failure modes are no                  !
l           different than the failure modes for the existing motor. The MOV has been determined by calculation to operate and achieve the thrust iequirements necessary to open and close the valve under both normal operating and accident conditions. Therefore, this change does not create the
        ' possibility of an accident or malfunction of equipment important to safety of a different type. The         )
l Technical Specifications make no direct reference to the minimum flow valve or its characteristics other than the operational status of the RHR system. This change maintains the valve functional requirements by upgrading the motor. The configuration has changed, however, there no changes

Attachment 2 to U-603276 Page 142 of 252 to the design basis, function or method of maintaining acceptable flow rates. Therefore, this change does not reduce a margin of safety as defined in the basis for any Technical Specification.
TEMPORARY POWER FEED TO RADWASTE BUILDING RISERS A 4
Activity Evaluated: Temporary Modification 98-075                             Log Number: 98-113          -
Due to maintenance work on 4160 volt bus l A (which feeds Unit Substation G), Temporary Modification 98-075 will be installed to provide temporary power to Radwaste Building Riser A
       ' from 480_ V Unit Substation J cubicle C. This temporary modification also requires the Motor
 ,       Control Center 0AP67E end its associated Heating, Ventilating, and Air Conditioning system be temporarily de-energized to allow the additional riser load to be added. This temporary modification only changes the source of non-safety power for a non-safety load of the Auxiliary
     , Power Sy:; tem. This temporary modification alters the electrical bus supply and loading configuration depicted on Updated Safety Analysis Report (USAR) Figure 8.3-1. Radwaste Building Riser A is not required to fimetion during any conditions other than normal station .
operating conditions. The new temporary source of power performs in the same function as the normal source. The Radwaste Building Riser operates in the same manner from either source.
All electrical bus loadings and short-circuit contributions are within equipment and cable ratings.
The components and equipment powered from the buses and risers are not credible initiators of any accident evaluated in USAR Chapter 15. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. Breaker tap settings provide overload protection as well as
       . coordination with upstream breakers. Since electrical loading is within ratings and the breakers are coordinated properly, all equipment failures would occur in the same manner as before with
   ,     similar consequences. Therefore, this change does not create the possibility of an accident or maltbaction of cquipment of a different type than previously evaluated. The load that this temporary modhication affects is non-safety related and does not impact any Technical Specification requirement. The equipment continues to perform its intended non-safety related ibnction. T herefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
RSP GETARS CIRCUITS 1RI85B AND 1RI85C ISOLATION .aM MCR DUE TO FIRE Activity Evaluated: ECN 30968                                                 Log Number: 98-114 Engineering Change Notice (ECN) 30968 uses existing spare Transfer Switch Sll isolation contacts in the IRI858 circuit and rewires circuit IRI85C from Terminal Board (TB) LL-19 and LL-20 to TB LL-7 and LL-8 to take advantage of existing isolation contacts. The wiring changes to the General Electric Transient Analysis Recording System (GETARS) circuits, IRI85B               j and 1RI85C provide isolation from potential " Fire Induced Circuit Damage" for the flow control             i input to the Reactor Core Isolation Cooling (RCIC) Turbine Ramp Generator and Signal                        l Converter and for the Tachometer output to the Remote Shutdown System RCIC speed indicator.

     <                                                                                        Attachment 2 to U-603276 Page 143 of 252               j
   ~
The Updated Safety Analysis Repon (USAR) does not identify any design basis accidents or                      j events involving GETARS. The_USAR does however evaluate an accident involving spurious operation of RCIC due to inadvertent RCIC pump startup. The two potential initiators for this
            . event, operator error and false start signal, are not affected by this design change. Therefore, this change does not increase the probability of an accident previously evaluated in the USAR Inadvertent initiation of RCIC system as previously evaluated in the USAR credits the trip logic           -
of the Main Turbine and Feedwater Pump Turbine to actuate within 20 to 25 seconds after RCIC                 I i
initiation, after which the event has the same consequences and operator actions as a Main-Turbine trip, except that RCIC is already nmning. This ECN only affects GETARS circuits                      ,
within the Remote Shutdown Panel (RSP) As such, it does not impact the Main Turbine or Feedwater Pump Turbine trip logics nor does it impact any operator actions credited for response
             ' to Main Turbir.e trip. Therefore, this change does not increase the consequences of an accident              j l-             previously evaluated in the USAR.' The GETARS system is only a data collection tocl used for                 )
post-transient response analysis and for data collection during certain Technical Specification              l surveillances: Any failure of GETARS would not adversely impact the RCIC system. The-rewiring involves short pieces of wire without changing the original function of the circuit, all of l           . the existing interlock functions and operational characteristics per original design are maintained.          I This activity ensures the capability of safe shutdown from the RSP without any possibility of" Fire Induced Circuit Damage" failure due to an accident fire in the Main Control Room. Therefore,
             ' this change does not increase the probability or consequences of a malfunction of equipment         34       j
             '. important to safety previously evaluated in the USAR. The rewiring of GETARS control circhitsig       '
maintain the same physical and electrical separation of divisional power sources as the original        '
circuitry and does not add any components or systems to the plant facility. Therefore, this change iloes not create the possibility of ari accident or malfunction of equipment important to safety of a        l
              'different type than previously evaluated. RCIC system operation and Remote Shutdown (RS)-
system operation per Technical Specifications are not affected by this change. GETARS is a data collection tool used during performance of some Technical Specification surveillances, but is not
              ~ itself a Technical Specification system. This ECN does not chhnge any operating parameter of the.
RCIC or RS systems. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REPLACE ACTUATOR ON RHR PUMP MINIMUM FLOW VALVE TO INCREASE TORQUE THR'UST -
                ' Activity Evaluatedi ECN 30933                                         Log Number 98-117 Engineering Change Notice (ECN) 30933 replaces the motor operator for motor-operated valve                 l;
              - IE12-F064B, Residual Heat Removal (RHR) Pump 1B Minimum Flow Valve, with a larger motor to incream actuator torque / thrust capability. The original assumptions for actuator                j
: efficiency used for actuator analyses, made by the valve operator manufacturer Limitorque, were             {
non-conservative; Limitorque Technical Update 98-01 and its supplement, reduced allowable actuator efficiency based on industry data. To compensate for the lower allowable efliciency of
;                the actuator, a larger motor is being installed to provide 10 foot pounds of torque as opposed to
[                the original 5 foot pounds.
4

Attachment 2               '
to U-603"?6 t
Page 144 of 252 Chapters 5,0, and 15 of the Updated Safety Analysis Report (USAR) discusses the RHR system for mitigation of any accident that ultimately results in a plant trip. The larger motor will not over Ltorque the valve internals when the torque switch settings are within the requirements. Also, the larger motor has no~ ad. verse affect upon the Motor Control Center'(MCC) bus loading or diesel generator loading, and is seismically acceptable. The motor change has no adverse affect upon
          ; the minimum flow valve or RHR system reliability, function, or method of performing its function.           -
: The minimum flow valve failure represents a single failure that has already been accounted for and, as a result, the dose consequences are not adversely affected; Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR.
When the flow in the main flow path falls to 750 gallons per minute, the minimum flow valve automatically opens and establishes a flow path to the suppression pool to prevent overheating and possible damage to the pump: The new motor has been procured to the same safety class as the original motor. 'This plant change does not interface with any other plant systems, and does not affect the design basis, function, or method of performing the functions of the RHR system.
Additionally, this plant change does not affect the design basis, function or reliability of the valve.
It is a configuration change to compensate for the lower allowable efficiency of the actuator. If the valve were postulated to fail open, the amount of coolant injected would be below the amount required in the USAR. This again constitutes a system single failure already credited in the -
accident analysis. If the valve were postulated to fait closed during pump operation, long term
  #         operation at low flow could damage the pump. Insta!!ation of the larger motor helps to ensure the M;gg@ valve performs its design basis safety-related function. Therefore, this chan
    & ' the probability or consequences of a malfunction of equipment imponant to safety previously
          . evaluated in the USAR. The failure modes are no different than the failure modes for the existing-motor. The MOV has been determined by calculation to operate and achieve the thmst .
requirements necessary to open and close the valve under both normal operating and accident conditions. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type. This change maintains the valve functional
          ! requirements by upgrading the motor. 'The cor. figuration has changed, however, there no changes to.the design basis, function or method of maintaining acceptable flow rates. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REPLACE ACTUATOR ON RHR PUMP MINIMUM FLOW VALVE TO INCREASE TORQUE THRUST                                                                                                 l Activity Evaluated: ECN 30934                                           Log Number: 98-118
          ' Engineering Change Notice (ECN) 30934 replaces the motor operator for motor-operated valve IE12-F064C, Residual Heat Removal (RHR) Pump 1C Minimum Flow Valve, with a larger motor to increase actuator torque / thrust capability. The original assumptions for actuator efficiency used for actuator analyses, made by the valve operator manufacturer Limitorque, were non-conservative. Limitorque Technical Update 98-01 and its supplement, reduced allowable actuator efficiency based on industry data. To compensate for the lower allowable efficiency of the actuator, a larger motor is being installed to provide 10 foot pounds of torque as opposed to
           ~t he original 5 foot pounds.
9

Attachment 2 m                                                                                   to U-603276 Page 145 of 252 Chapters 5,6, and 15 of the Updated Safety Analysir Leport (USAR) discusses the RHR system for mitigation of any accident that ultimately results in a plant trip. The larger motor will not over torque the valve internals when the torque switch settings are within the requirements; Also, the larger motor has no adverse affect upon the Motor Control Center (MCC) bus loading or diesel
         - generator loading, and is seismically acceptable. The motor change has no adverse affect upon the minimum flow valve or RHR system reliability, function, or method of performing its function.        .
         ~ The minimum flow valve failure represents a single failure that has already been accounted for
         - and, as a result, the dose consequences are not adversely affected. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR.
When the flow in the main flow path falls to 750 gallons per minute, the minimum flow valve automatically opens and establishes a flow path to the suppression pool to prevent overheating
          . and possible damage to the pump. The new motor has been procured to the same safety class as the original motor.; This plant change does not interface with any other plant systems, and does .
                                                      ~
          . not affect the design basis, function, or method of performing the functions of the RHR system.
Additionally, this plant change does not affect the design basis, function or reliability of the valve. t It is a configuration change to compensate for the lower allowable efficiency of the actuator. If           {
the valve were postulated to' fail open, the amount of coolant injected would be below the amount
: required in the US AR. This again constitutes a system single failure already credited in the adcident analysis. If the valve were postulated to fail closed during pump operation, long term o            operation at low flow could damage the pump. Installation of the larger motor helps to ensure the valve performs its design basis safety-related function. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The failure modes are no different than the failure modes' for the existing motor.L The MOV has been determined by calculation to operate and achieve the thmst%
requirements necessary to open and close the valve under both nonnal operating and accident conditions. Therefore, this change does not create the possibility of an accident or malfunction of l
i            ' equipment important to safety of a different type. This change maintains the valve functional l             requirements by upgrading the motor. The configuration has changed, however, there no changes to the design basis, function or method of maintaining acceptable flow rates. Therefore, this-l activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
L             REPLACEMENT OF DISPLAY CONTROL SYSTEM CRT MONITORS WITH TOUCH l:     -    ' SCREEN MONITORS-Activity Evaluatedi ECN 30901; USAR Change 8-212                                Log Number: 98-119 Engineering Change Notice (ECN) 30901 replaces Cathode Ray Tube (CRT) monitors, used by the Display Control System (DCS), with infrared touch-screen displays. The new touch-screen
            ' displays are comparable in size, display, and function to the CRT displays. The new displays are powered from the same power sources as the CRTs with new surge suppressors added for
             - enhanced protection ~of the touch screen unit. The DCS is used by the operators and emergency
            . response personnel to monitor plant conditions ander normal and emergency conditions.
However, no credit is taken for this system to perform safe shutdown of the plant in accident conditions and t' he DCS is'not credited with mitigation of any accidents or transients discussed in i

Attachment 2 j                                                                                                 to U-603276
!-                                                                                                Page 146 of 252
: the Updated Safety Analysisleport (USAR). The DCS system is not safety related, not Seismic Category 1, non-Class IE, and does not control any systems in the plant. The infrared touch screen displays are designed, manufactured, installed, and tested in accordance with industry codes and standards,' USAR requirements, and applicable plant specifications in order to maintaic p                  system reliability Also, due to separation and independence appropriate fuse and breaker coordination, a failure in the DCS will not initiate any evaluated transient or accident. The total       -
electrical, thermal, and seismic loading of the new touch-screens is equivalent to the total loading of the CRTs; Therefore, this change doe:: not increase the probability or consequences of an
               - accident or malfunction of equipment impodant to safety previously evaluated in the USAR. The Display Control System does not control or operate any systems in the plant. There are no new
               - system interactions associated with this change, and the system will continue to function as
: before; Therefore, this change does not create the possibility of an accident or malfunction of
                  . equipment important to safety of a different type than previously evaluated in the USAR. The Technical Specifications do not address the Display Control System, nor is the system credited with providing a margin of safety supporting any other Technical Specif; cation requirement.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
YEAR 2000 HARDWARE / SOFTWARE CHANGE TO PROCESS MONITORING SYSTEM Activity Evaluated: Modification CX-032; USAR Change 8-216                     Log Number: 98-120 Modification CX-032 upgrades the Process Monitoring System (CX) to provide capability to deal
               - with thyYear 2000 problem facing computer based systems. The CX system is used by the
                  ' operators and emergency response personnel to monitor plant conditions under normal and
               - emergency conditio~ns, but there is no credit taken for the availability of the CX system in the accidents and transients analyses in the Updated Safety Analysis Report (USAR). The CX system is not safety related, not Seismic Category I, non-Class 1E, does not control any systems in the
               ' plant, and is not required for safe shutdown of the plant. The CX system is designed with appropriate separation and independence from the systems important to safety to preclude any
                  . adverse affects that could cause any of the accidents and transients evaluated in the USAR.
Neither the software, nor the hardware changes alter the overall function of the CX system. The changes are verified and validated by the vendor to correct the problems identified and not cause any' adverse affects on the operation of the system. Addition of the Data Memory Address (DMA) printed wiring boards provide an interface with a new computer which provides an
                ' alternate data acquisition interface for the CX system. The additional power loading of the new DMA boards was determined to have minimal impact on the existing power supplies associated with the Nuclear Steam Supply and Balance of Plant processors. The hardware and software
                ' changes are designed, manufactured, installed, and tested in accordance with industry codes and standards, USAR requirements, and applicable plant specifications in order to maintain system reliability. Therefore, this change does not increase the probability or consequences of an accident or malfunction ~of equipment imponant to safety previously evaluated in the USAR. The Process Monitoring System does not control or operate any systems in the plant. There are no new system interactions associated with this change, and the system will continue to function as            i

r 4 ,
l    w                                                                                               Attachment 2 to U-603276           )
Page 147 of 252
(:
1 before. Therefore, this change does'not create the possibility of an accident or malfunction of
             . equipment important to safety of a different type than previously evaluated in the_USAR. The Technical Specifications do not address the Process Monitoring System, nor is the system L            : credited with providing a margin of safety supporting any other Technical Specification
             - requireraent. Therefore, this activity does not ieduce a margin of safety as defined in the basis for p            ' any Technical Specification.                                                                              -
DISABLING SHUTDOWN SERVICE WATER SYSTEM LOW PRESSURE SWITCH TRIP L FUNCTION FOR CONTROL ROOM CHILLERS
                                                               ~'
              ; Activity EvaluatedL ECNs 31081 & 31087; USAR Change 8-226 '                          Log Number: 98-122-l Engineering Change Notices (ECNs) 31081 and 31087 disable pressure switches associated with the Control Room (VC) chillers, VCl3CA/B. These pressure switches alarm and trip VC chillers-upon low Shutdown Service Water (SX) line pressure. VC chillers provide chilled water to the               ;
cooling coils in each air-handling equipment train from one of two separate water circuits from
             . the control room chilled water system. . Deletion of the trip function is not an equipment protection concern; because, in the event oflow SX pressure, the SX heat sink capabilities will be diminished and result in higher refrigerant pressure and temperature. Condenser high refrigerant
_ pressure switches and high temperature switches are available to trip the chillers based on high refrigerant pressure or temperature signal. According to Updated Safety Analycis Report                    ]
(USAR) Section 9.2.8, the control room chilled water system is required to be operable.during a seismic event and a Loss of Offsite Power (LOOP). There are no credible railure modes due to               1 removal ofinterlock that can act as an initiating event for any evaluatca accident during a seismic        l event or LOOP, Removal ofinterlock does not cause operation of chiller outside design L parameters, sinimizes the potential for an unnecessary trip due to pressure switch spurious
!               actuation, thus increasing availability / reliability of the chiller, This activity does not degrade any system or component performancer This activity does not impact design, material, or constmetion standards applicable to the control room chiller design. Therefore, this change does           .
not increase the probability or consequences of an accident or malfunction of equipment
              ' important to safety previously evaluated in the USAR. Original design of the system would trip -
the VC chillers upon loss or reduction of SX pressure. Since the high refrigerant pressure and high temperature trips are available, the removal of the interlock does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The VC system is designed to provide a radiological controlled environ' ment to ensure habitability of the control room for the safety of control room operators J under all plant conditions ~. - Two independent trains are each capable of performing the stated
              ' function? The instrument and controls automatically initiate action to route make-up air to Main Control Room through emergency filters.- Those instruments are not directly impacted or altered L               by'this activity, and thus, this activity does not affect acceptance values for the system, structure,     !
l'         _
component, or function identified in the Technical Specifications. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
l ';
l L                                                                                                                          :

E Attachment 2           '
to U-603276 -
Page 148 of 252 TS BASES 3.8.2 CLARIFICATION FOR SR 3.8.2.1 Activity Evaluated: TS Bases Change BL-98-017.                                   Log Number: 98-126 Note: These changes superceded by Safety Evaluation Log Number 98-130.
Technical Specifications (TS) Surveillance Requirement (SR) 3.8.2.1 identifies the surveillance tests that apply'to the diesel generators (DGs) during shutdown conditions. The applicable tests are identified by listing what SRs from TS 3.8.1, "AC Sources - Operating," are applicable under
       - TS 3.8.2,"AC Sources - Shutdown". SR 3.8.2.1 contains a Note which allows cenain SRs not to
       . be met during s' hutdown conditions. The Bases for SR 3.8.2.1 provides the basis for the Note.;In response'to an identified discrepency between the Note and its basis, this activity revises the Bases to more clearly convey the justification and applicability of the Note. . Initiating events for
  , ' . accidents or transients while shut down include fuel handling and operations with a potential for draining the reactor vessel. Failure of any DG itselfis not an initiator of these postulated accidents. The change does not create any system performance issues for the DGs including their supponing or supported systems. There is no change to the DGs or the requirements for the number of OPERABLE DGs or to the surveillance requirements to declare a DG OPERABLE.
Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety _ Analysis Report (USAR). This activity does not involve changes to the test requirements themselves, it
       ' only affects what testing requirements'must be met. The change does not introduce any new modes _of testing or operation, nor does it involve any design changes or physical changes to the facility. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated, This activity does not involve any changes to the acceptance values, design limitations, safety limits, or Limiting-
       ~ Conditions for Operation specified in the Technical Specifications. Therefore, this Technical Specification Bases change does not reduce a margin of safety as defined in the basis for any Technical Specification.
FLOW RESTRICTION DEVICE IN THE BACKWASH LINE FOR DIVISION 2 SHUTDOWN SERVICE WATER STRAINER 1 Activity Evaluated: ECN 31102; USAR Change'8-209                                 Log Number; 98-l'27 Engineering Change Notice (ECN) 31102 installs a flow restriction orifice in the backwash line ISX29BB-3 for Division 2 Shutdown Service Water (SX) system strainer ISX01FB, The orifice will be installed downstream of the existing motor operated isolation valve ISX013E. The SX system provides a reliable source of cooling water for station auxiliaries essential for a safe reactor shutdown following a Design Basis Accident (DBA). Based on calculation
                                  ~
IP-M-0486 R/2, it was determined that with the backwash line in-service the SX system c.an operate within its design limits in all modes requiring SX services. Failure of the orifice to allow adequate backwash flow may accelerate strainer plugging. The differential pressure drop across the strainer will increase until alarm point. This condition is enveloped by the existing analysis in

7 c-
          .o                ,
Attachment 2 to U-603276 Page 149 of 252
                . Updated Safety Analysis Report (USAR) Table 9.2-6. Therefore, this change does not increase
               ' the probability of an accident previously evaluated in the USAR. The activity ensures design flows are met throughou'tthe SX system components so that the primary and secondary                       i
                ! containment can perform its intended function. This activity does not change the SX system design flows, its modes of operation, nor does it change, degrade, or prevent actions described or assumed in mitigating the DBA~ Therefore, this change does not increase the consequences of an          .
accident previously evaluated in the USAR. In the unlikely event the orifice fails to allow backwash due to obstruction, this malfunction will not impair the operation of the SX system.
This activity does not change the heat removal or design function of the system. With the SX system operating per the original design requirements, the interacting system which are dependent of the SX flow'are expected to perform as designed. Therefore, this change does not increase the
                - probability or consequences of a malfunction of equipment important to safety previousiv evaluated in the USAR. Since, the SX system is an accident mitigation type system and the flow
    ,             restriction orifice is a passive device; this change does not create an accident or equipment
                                                                                          ~
malfunction of a different type. The orifice serves to modify the SX system such that the design q           basis is maintained in accordance with the USAR. This change assures that flows through other SX components meet design requirements whether or not the strainer backwash is in service. The Technical Specifications _for the SX . system are unaffected because the functional characteristics of the system are unchanged.~ Therefore,' this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
ELIMINATION OF REQUIREMENT TO PERFORM A CHANNEL FUNCTIONAL TEST OF THE TEMPERATURE INSTRUMENTATION OF THE MSIVLCS AT LEAST ONCE PER 31 DAYS
: Activity Evaluated: ORM Change 22-8                                             Log Numberf 98-128 Operational Requirements Manual (ORM) Section 4.4.7.2 requires a Channel Functional Test of the Main Steam Isolation Valves Leakage Control System (MSIVLCS) flow, pressure and temperature instrumentation at least once per 31 days. ORM Change 22-8 deletes the requirement to perform a Channel Functional Test on the temperature instrumentation of the z MS1VLCS at least once per 31 days. The temperature in;trumentation is only found on the Inboard MSIVLCS system and has no control, permissive, interlock, alarm or trip functions. The
                  -temperature instmmentation only serves to indicate the temperature ofinboard MSIV leakage flow out'of the electrically heated pipe sections of the system. This activity does not compromise
                 ' the design, material, or construction standards to which the MSIVLCS was constructed and is being maintained.i Temperature in'strumentation is. not used to control the MSIVLCS process and does not contribute to the mitigatory function of the MSIVLCS. Also, this activity does not -
introduce any components or change any components that.would cause the MSIVLCS to be                       I
                 . operated outside ofits' design or testing limits. The heater operation will still be verified once every 31 days per Technical Specification Surveillance Requirement 3.6.1.8.2 by verifying heater        j element circuitry. Also, the temperature indication will be calibrated every 18 months per ORM          ;
4.4.7.3 to ensure temperature indication is operable and accurate. This activity does not impose
                  ; increased or more severe testing requirements, nor does it change any saismic, environmental or        )
l l
4 I
l

Attachment 2 -
  ^                                                                                              to U-603276 Page 150 of 252 separation criteria. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment importam to safety previously evaluated in the Updated
                                                                        ~
: Safety Analysis Report (USAR). This ORM change does not change any component or system in th'eplant, nor does it introduce any component or system into the plant. The MSIVLCS is still able'to perform its safety function independent of whether its temperature instmmentation is
: operable or not. Therefore,~ this ' activity does not create the possibility of an accident or ~          -
             - malfunction of equipment important to safety of a different type than previously evaluated.
Technical Specification 3.6.1.8.3 discusses operability requirements for the MSIVLCS and for the MSIVLCS subsystems and components. This activity does not impair the MSIVLCS operability or prevent the MSIVLCS from performing its safety function. Therefore, this ORM change does not reduce a margin of safety as defined in the basis for any Technical Specification.
             . CONNECTION OF A TEMPORARY BATTERY CHARGER TO THE BOP (NON-SAFETY)
              -DC BUSES Activity Evaluated: Temporary Modification 98-080                             Log Number: 98-129 R/l L Temporary Modification 98-080 installs a battery charger with a 300 Amp rated output to maintain the Balance of Plant (BOP) DC I F Bus, DC Motor Control Center (MCC) IF Distribution Panel, and DC MCC IE Distribution Panel energized and the BOP battery IF on float charge. This temporary modification was only in place for the time needed to perform maintenance on the BOP battery chargers and was removed prior to start-up. Updated Safety
: Analysis Report (USAR) Chapters 6 and 15 do not discuss any accident scenarios associated with the non-divisional DC buses. Loss of these buses can increase the burden on the operators to
: monitor and control the plant, but they are not required for the response to any accident. This 4
temporary modification adds combustibles to the areas in which the temporary cables are installed, but these two cables and the battery charger do not represent enough fire load to cause a change              !
in the fire load rating of the areas involved. The in.stallation was performed in accordance with site procedures. The operation of the DC BOP system is constrained due to the reduced output capability of the connected temporary charger. The only interaction between the non-safety DC buses and the safety related equipment is the Reactor Protection System (RPS) inverters which
: are fed from the non-safety DC system. The load represented by the. RPS inverters is unchanged by the temporary modification installation and is within the rated output of the temporary battery charger, The charger installed by this temporary modification performs with the same level of reliability as the standard charger and does not introduce a new failure mechar. ism into the system.
Therefore, this change does not increase the probability or consequences of an accident or-malfunction of equipment important to safety previously evaluated in the USAR. The electrical                  ;
source used to feed the temporary modification wa's reviewed to confirm its capability to carry the            !
additional load 'of the charger. This temporary rrodification could cause loss of non-safety related           j computers, that would create an increased burden on the operators to monitor plant parameters;                 I I
howeverlthis is acceptable since the parameters important to ensure the integrity of the radiation boundaries are monitored with safety related instrumentation. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated / Technical Specification 3.8.7 discusses RPS inverters and their e-a

   .                                                                                      Attachment 2 to U-603276 -
Page 151 of 252 preferred feed from the DC source. Thit feed is not impacted by this temporary modification.
  ~
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
TS BASES 3.8.'2 CLARIFICATION OF SR 3.8.2.1 Activity Evaluated:' TS Bases Change BL-98-017                                Log Number: 98-130 Technical Specifications (TS) Surveillance Requirement (SR) 3.8.2.1 identifies the surveillance tests that apply to the diesel generators (DGs) during shutdown conditions._ The applicable test i        are identified by listing what SRs from TS 3.8.1, "AC Sources - Operating," are applicable under TS 3.8.2, "AC So~urces - Shutdown". : SR 3.8.2.1 contains a Note which allows certain SRs not to be met during shutdown conditions. The Bases for SR 3,8.2.1 provides the basis for the Note. In response to an identified discrepancy between the Note and its basis, this activity revises the Bases to more clearly convey thejustification and applicability of the Note. Initiating events for -
accidents or transients while shut down include fuel handling and operations with a potential for draining the reactor vessel. Failure of any DG itselfis not an initiator of these postulated accidentsc The change does not create any system performance issues for the DGs including their supporting or supported systems. There is no change to the DGs or the requirements for the number of Operable DGs or to the surveillance requirements to declare a DG Operable.
Therefore, this activity does not increase the probability or consequences of an accident or    ~
malfunction of equipment important to safety previously evaluated in the Updated Safet'y Analysis Report (USAR). ' This activity does not involve changes to the test requirements themselves, it only affects what testing requirements must be ~ met. The change does not introduce any new          ,
modes of testing or operation, nor does it involve any design changes or physical changes to the -
facility. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This activity does not involve any changes to the acceptance values, design limitations, safety limits, or Limiting Conditions for Operation specified in the Technical Specifications. Therefore, this Technical Specification Bases change does not reduce a margin of safety as defined in the basis for any
_ Technical Specification.
OPERATIO'NOF RHR HX WITH THE SERVICE WATER INLET AND OUTLET VALVES OPEN
       . l Activity E0aluated: CR 1-98-07-105                                             Log Number: 98-134 Condition Report (CR) 1-98-07-105 was initiated stating that a change made via a safety-screening had failed to determine that the activity was a change to procedures as described in the Updated Safety Analysis Report (USAR). This Safety Evaluation was performed to demonstrate
         - that an unreviewed safety question was not involved. This activity changes the positions for the Residual Heat Removal (RHR) A/B Heat Exchangers (HXs) Shutdown Service Water (SX) inlet and outlet isolation valves to allow flow through the HXs in normal standby lineup. The line-up

Attachment 2 to 'U-603276 Page 152 of 252           i of service water through the RHR HX in' normal operation will not be permitted nhen the temperature of the service water is 43*F or less, or in operating conditions other than Mode 4.
                                                                                                                    ]
The following accidents potentially involve the RHR Heat Exchanger: loss of one of the two RHR Heat Exchangers, inadvertent operation of RHR shutdown cooling, loss of shutdown cooling, and decrease in reactor coolant inventory. Since this activity will be performed during        -
      . Mode 4 with SX verified to be continuously flowing through the RHR HX, it does not increase
      - the probability of these accidents to' occur, The RHR H.Xs and SX system are used to mitigate the. consequences of many of the accidents evaluated in the Updated Safety Analysis Report
      - (USAR)J None of the equipment required to function will be in an unanalyzed condition, nor.will
         'there be the possibility of exceeding any operating limits or setpoints. This configuration allows the Plant Service Water pumps to continue'to supply the SX system loads, including the
: RHR HXs as originally . designed. The continuous flow of SX through the TUIR HX room can
         . serve asan additional heat sink for the RHR HX room.' However, this valve line-up will not be allowed when SX water temperature is at or below 43 F. Thus, this bhange cannot contribute to
      - the RHR HX room dropping below the minimum normal temperature of 40 F. Also, this activity-is limited to a lake temperature of greater than 43 F since the RHR shell side of the RHR heat
: exchanger is designed for a temperature range of 40 to 480 F, Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This activity does not add any new structures, systems or. components, or change any system interactions. This activity is consistent with system design and provisions for operation. Therefore, this activity does not create the possibility of an accident
       - or malfunction of equipment important to safety of a different type than previously evaluated in j     the USAR. This activity does not affect any Technical Specification safety limit, limiting safety system setting, or limiting condition for operation. In addition, Technical Specification requirements for SX and RHR Operability.will continue to be fully met. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specificatio'n.
      ' EMERGENCY OPERATING PROCEDURES / SEVERE ACCIDENT GUIDELINES .
PROGRAM AND PROCEDURES Activity Evaluated: Implementation of the BWROG                                ' Log Number: 98-135 Emergency Precedure Guidelines and Severe Accident Guidelines Rev.1 This activity implements Revision 1 of the Boiler Water Reactor Owners' Group (BWROG)
Emergency Procedure and Severe Accident Guidelines (EPGs/ SAGS). These EPGs/ SAGS provide generic symptomatic direction for BWR emergency response and severe accident mitigation. The EPG/ SAGS act as an accident mitigation aide in response to Design Basis
: Accidents (DBAs); thus, this activity cannot act as an accident initiator in any of the DBAs
      ' defined in Chapters 6 and 15. These EPGs/ SAGS are not procedures to operate any equipment during normal operation and would be used after a failure has already occurred or systems and components are no longer responding as designed. The conditions addressed by the EPGs/ SAGS are beyond plant design basis conditions. Therefore, this activity does not increase the probability

h             4         .
L
         -                                                                                            Attachment 2 to U-603276 Page 153 of 252 (L          ,
or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). Nor does this activity create an accident or malfunction of equipment important to safety of a different type than previously
                , evaluated in the USAR. This activity does not cha~nge any Techni cal Specification safety limits,          '
L limiting safety system settings, limiting conditions for operat' ion, or the bases for these
                 - requirements. Therefore,' this activity does not reduce a margin of safety as defined in the basis ;      . i l
for any Technical Specification.                                                                            i
          ,          CORRECTIVE ACTION TO REDUCE VOLTAGE LOSS IN DIVISION III DG OUTPUT BREAKER CONTROL CIRCUIT l.
Activity Evaluatedi ECN 31064:                                                   Log Number:' 98-136
   ^
Engineering Change Notice (ECN) 31064 rewires the closing control circuit for the Division III High Pressure Core Spray (HPCS) system Diesel Generator (DG) output Meaker. The purpose                     '
e
                 ' of this activity is to maintain the existing circuit logic,while reducing the overall wiring L                    impedance. !The work is all internal to the panels and some internal jumpers were added and l                - removed; The HPCS system is an accident initiator for a coolant inventory increase due to l                    inadvertent actuation of the HPCS pump. This scenario assumes an operator error or false start signal for the pump. The power'' source is not discussed but would have to be available for the
[
pump to operate. :The DG power source.will not initiate the event since the control circuits for-the pump and DG breakers are distinct and separate. Since, this ECN does not change the logic              ,
of the DG output breaker control circuit nor the operator controls or interface, this change does          1 not increase the probability or consequences of an accident previously evaluated in the Updated l                   l Safety Analysis Report (USAR). The HPCS system functions to maintain coolant inventory while
!.                 . t' he vessel is pressurized. By assuring adequate voltage is applied to the closing coil for the DG L                    ' output breaker, the activity supports the expectation that the on-site power source is able to close onto the bus when requir'ed to do so. Since the design change does not introduce any new logic L
                 . or affect how the system responds to any ofits initiation signals, this activity does not increase the probability of a malfunction of equipment important to safety previously evaluated in the USAR.
L                                                                                                                                ,
L                     The work performed under this activity uses existing spare conductors and some new internal                j paneljumpers. _ These conductors and the termination hardware used to land them meet the same              ;
L                     design standards and quality as the material used in the original. circuit. By reducing the voltage        !
drop in the circuit, the design assures that the voltage supplied to the operating coil is above the minimum necessary. Therefore; this change does not create the possibility of an accident or a L                 . malfunction of equipment important to safety of a different type. The Technical Specifications do            .
i l_                   . not' discuss the operating voltage of the control coils for the breakers involved only that the breakers will operate when required. Other specifications pertaining to the HPCS system are
                    ~~ unaffected by virtue of there being no change in circuit logic or functional performance.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical     l
                    - Specification.                                                 ,

Attachment 2 to U-603276 Page 154 of 252 CLARIFICATION TO USAR SECTION 9.3.7.2 AND APPENDIX D SECTION II.B.3-1 Activity Evaluated: USAR Change'8-218 :                                        LogNumber: 98J137 Updated Safety Analysis Report (USAR) Change Package '8-218 revises Sections 9.3.7.2 and Appendix D, Section II.B.3-1 to clarify that "Off-line radionuclide analysis can be performed on
     ~ diluted samples of: liquid, drywell atmosphere and containment atmosphere." These changes do not change the implied _ meaning of the paragraph and do not contradict any other section of the
: USAR. This activity does not clange the functions or capabilities of any Post Accident Sampling System (PASS) equipment. The activity clarifies the sampling capabilities of the PASS as described in the vendor's design documentation. Failure of the system will not compromise any safety-related systems or prevent safe shutdown of the plant. In addition, this activity does not affect the operability, availability, or reliability of the PASS equipment. Also, the clarification of
    - the PASS sampling capabilities does not change, degrade, or prevent the system from performing its function to estimate core damage. This system does not provide a mitigatory function in any USAR evaluated accident. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This USAR change is a clarification only, it does not introduce or change any component in the plant. Therefore, this activity does not create the possibility of an accident or
                                                 ~
malfunction of equipment important to safety of a different type. Technical Specification 5.5.3 describes the requirements of the Post Accident Sampling System. The requirements include a -
program that will p~r ovide controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident coaditions. These requirements are not affected by the editorial change to the USAR; Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
EMERGENCY PLAN MINIMUM STAFFING REQUIREMENTS
     ' Activity Evaluated: CPS E-Plan Rev/l1, USAR Change 8-215,                     Log Number: 98-138 and ORM Change 22-9 The Clinton Power Station (CPS) Emergency Plan (E-Plan) is being revised to more accurately reflect the on-shift and augmented resources needed to respond to an accident and protect the public health and safety. This change to the Emergency Plan will enhance the response to the
     ~ design basis accidents (DBA) discussed in the Updated Safety Analysis Repon (USAR) by ensuring adequate stalling is available on shift and additional personnel are available in                i approximately 60 minutes. Emergency Plans are desiped to react to accidents, not create new means of an accident. This change does not have an affect on any plant design, system, structure, or compo'nent. A DBA is a non-mechanistic accident that is not affected by the augmentation of             ;
the Emergency Response Organization._ Actions taken by the E-Plan are independent of actions taken by operators in the manipulation of plant equipment. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to          j safety previously evaluated in the USAR. This change to the CPS E-Plan deals with staff i

y    .     .
_.,                                                                                          Attachment 2 to U-603276 Page 155 of 252
               ~
             . augmentation and is designed to enhance'the response to any accident, it does not affect plant equipment or the operation of plant equipment. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety of a different type.
The Technical Specifications do not establish limits associated with the Emergency Plan other than requirements for the plan, its implementing procedures, and periodic reviews by the review and audit organizationsi This change does not alter these requirements or degrade the ability of         -
the review and audit organizations to perform this function. ~ Therefore, this activity does not -
reduce a margin of safety as defined in the basis for any Technical Specification.
             . FL,0W RESTRICTION DEVICE IN THE BACKWASH LINE ISX29BA-3 FOR STRAINER 1SX01FA'- DIVISION 1 SX
             ' Activity Evalua't ed: ECN 31092; USAR Change 8-217                                    Log Number: 98-140 i
Engineering Change Notice (ECN) 31092 installs a flow restriction orifice in the backwash line ISX29BA-3 fcir Division 1 Shutdown Service Water (SX) system strainer ISX01FA. The orifice will be installed downstream of the existing motor operated isolation valve ISX013D. The SX system provides a reliable source of cooling water for station auxiliaries essential for a safe
: reactor shutdown following a Design Basis Accident (DBA). Based on calculation
              - IP-M-0486 R/2, it was determined that with the backwash line in-service (i.e., consistent with current design and operation basis), the SX system can operate within its design limits in all modes requiring SX services. Failure of the orifice to allow adequate backwash flow may accelerate strainer plugging. The differential pressure drop across the strainer will increase until an alarm setpoint is reached. This condition is enveloped by the existing analysis in Updated Safety Analysis Report (USAR) Table 9.2-6 Therefore, this change does not increase the probability of an accident previously evaluated in the USAR. The activity ensures design flows are met throughout the SX system components so that supported systems, structures, and components may perform their function. This activity does not change the SX system design flows,- its modes of operation, nor does it change, degrade, or prevent actions described or.
as'sumed in mitigating the DB A Therefore, this change does not increase.the consequences of an
              . accident previously evaluated in the USAR: In the unlikely event the orifice fails to allow backwash due to obstruction, this malfunction will not impair the operation of the SX system.
This activity'does not change the heat removal or design function of the system. Therefore, this         ,
change does_'not increase the'probabilitp or consequences of a malfunction of equipment
              - important to safety previously evaluated in the 'USAR. Since, the SX system is an accident mitigation type system and the flow restriction orifice is a passive device; this change does not create an accident or equipment malfunction of a different type. The orifice serves to modify the        l SX system such that the design basis is maintained in accordance with the USAR. This change assures that flows through other SX comp,onents meet design requirements whether or not the strainer backwash is in service. This change and'its preservation of assumed system flow rates preserves the initial conditions assumed _for the SX system in the Improved Technical
              . Specifications; Therefore, this activity does not reduce a margin of safety as defined in the basis i
: for anyTechnical Specification.
l 9

Attachment 2 to U-603276 Page 156 of 252 CHANGE ACTUATOR MOTOR FOR SX ISOLATION VALVE ISX014B TO INCREASE
    . TORQUE THRUST
    ~ Activity Evaluated: ECN 31174'                                                 Log Number: 98-145 Engineering Change Notice (ECN) 31174 replaces the motor in the actuator for motor operated valve (MOV) ISX014B, Shutdown Service Water System (SX) Isolation Valve. The new motor is of the same design as the existing motor; except that it: develops 5 ft Ib torque, rates at 0.33 horsepower, has an EB48 frame, and utilizes Type RH insulation.' The Shutdown Service Water (SX) system provides a reliable source of cooling water for several safety related components which are essential for a safe shutdown of the reactor following a Design Basis Accident (DBA).
The higher torque provided by the new actuator motor meets the design basis safety-related function torque / thrust at system Degraded Voltage (DV) in accordance with Limitorque
     ' Technical Update (TU) 98-01.
The SX system is an accident mitigation system. The Updated Safety Analysis Report (USAR) does not evaluate any accidents that are initiated by SX. This design change ensures the MOV actuator capability meet the safety related functional requirements. Therefore, this change does not increase the probability _ or consequences of an accident previously evaluated in the USAR.
Consistent with the existing actuator configuration, the larger motor will not over torque the valve internals when the torque switch settings are within the requirements. Also, the larger motor has no adverse affect upon the Motor Control Center (MCC) bus loading or diesel
    . generator loading and is seismically acceptable. The larger motor requires more current during opening and closing the valve, calculations in the ECN confirm that there is no adverse impact on the electrical system by the addition of the larger motor. Therefore, this change does not increase the probability of a malfunction of equipment important to safety evaluated previously in the USAR. These design changes improve the performance of the valve / actuator assembly for postulated conditions and do not degrade the performance of the valve and its capability to provide isolation. Therefore, this change does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This design change replaces an actuator motor with another motor of the same design and type, there are no new components being introduced into the plant. The failure modes are no different than the failure modes for the existing motor. The MOV has been determined by calculation to operate and achieve the thrust requirements necessary to open and close the valve under both normal operating and accident conditions. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type. This change maintains the valve functional
     . requirements by upgrading the motor. The configuration has changed, however, there no changes to the acceptance limits, design basis, function or method cf maintaining acceptable flow rates.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.-
                                                                                                                 )

rs
      .                                                                                          Attachment 2 L                                                                                                1o U-603276
        ^
;                                                                                                Page 157 of 252 i
RR SYSTEM PLC CPU REPLACEMENT -
              - Activity Evaluated: ECNs 31129/31130 -                                           Log Number: 98-146 Engineering Change Notices (ECNs) 31129 and 31130 involve replacing the Central Processing l               Unit (CPU) boards' located in the Reactor Recirculation (RR) Flow Control Valves' (FCVs)
L               Programmable Logic Controllers (PLCs), with a new CPU board. The original software in the PLC CPU is downloaded to an external computer for storage while the CPU board change is l             ' performed.; This activity does not require any other changes such as wiring changes, hardware l             : changes, or software changes. These changes are being implemented to correct Year 2000 L               problems with computer based systems.
The PLCs are required to lock the FCVs in an "as is" position if the stroking rate of the FCVs .~
exceeds 11% of full stroke per second. The RR HPU PLC can fail such that the RR FCV strokes at its maximum hydraulically limited rate. This may result in either an increasing or decreasing recirculation flow transient. The replacem'ent of the CPU boards does not affect the overall L             . function of the PLC since the existing programming, and logic are not changed. The replacement '
CPU boards are designed, manufactured, verified, validated, installed and tested to the applicable requirements'of the Updated Safety Analysis Report (USAR), industry codes and standards, and l
the manufacturer's requirements to ensure the reliability and adequate performance of the PLCs.
s    Therefore, this change does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the USAR. ! Radioactivity is discharged in a I
RR FCV Failure-Decreasing Flow Event. This accident analysis assumes that the FCV will stroke at its maximum rate which is limited bp hydraulic means. A failure of the RR PLC CPU can not cause a FCV to stroke faster than the maximum hydraulically restricted rate, and therefore, can not increase the c~onsequences of an accident previously evaluated in the USAR.~ This activity i              ~ does not create any new interactions with other systems important to safety or any new failure modes. Therefore, this change does not create an accident or malfunction of equipment important to.' safety of a different type. Technical Specification (TS) Bases 3.4.2 discusses the assumption that the FCV stroke rate is hydraulically limited to s 30% per second in the opening direction and s; 60* Aper second in the closing direction on a control signal failure of maximum demand. These stroke rates are assumptions of the accident analysis of the recirculation flow control failures for recirculation decreasing and increasing flows.' Replacement of the HPU PLC CPU board does not impact the hydraulic' limits discussed in the TS Bases. Therefore, this activity does not reduce ~a margin of safety as defmed in the basis for any Technical Specification.
CLARIFY RADIATION ZONE C DESIGN AND CONSTRUCTION CONSIDERATIONS                                        I Activity Evaluated: USAR Change 8-222                                           Log Number: 98-148 I          '. 'A Zone C radiation area is 2.5 to 20 millirem / hour and encompasses many of the plant accessible areas that contain equipment that might and will require manual operation, inspection, or maintenance during unit operation. The last sentence in Updated Safety Analysis Report (USAR)
               . Section 12.3.1.1, subparagraph c, states: " Auxiliary equipment requiring manual operation, f
r

o Attachment 2 to U-603276 -
Page 158 of 252 l
inspection, or maintenance during unit operation is not located in Zone C areas." This statement            j should be clarified because it is contrary to the actual operating zones that are determined by.            )
periodic radiation measurements. The sentence is revised to state: "Whenever possible, based on l
design and construction considerations, auxiliary equipment requiring manual operation,                      '
inspection,~ or maintenance during unit operation is not locate'd in Zone C areas." This change does not involve. operation, maintenance, or design of equipment important to safety, nor does it
    ' affect the environment of that equipment. Clinton Power Station shielding is designed to allow              i personnel access at a frequency and duration that is expected post-accident so that dose in vital areas'does not exceed the guidelines _ of General Design Criteria 19 of 10CFR50 (Code of Federal Regulations),' Appendix A.' In accident situations where exposures may exceed 10CFR20 limits _to prevent large releases of radioactivity to the public, implementation of applicable Emergency Plan Implementing Procedures will provide the necessary guidance in controlling the exposure of emergency personnel. This activity does not degrade or prevent equipment from performing actions described or assumed in the accident analysis, nor does it affect any system important to safety such as to impede access to vital areas. This activity does not introduce any new failure modes for equipment important to safety, nor does it introduce any new system, stmeture, or componentJ Therefore, this change does not create an accident or equipment malfunction of a different type than previously evaluated in the IJpdated Safety Analysis Report (USAR). No systems or equipment acceptance limits are impacted by this change. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
ACCESS TO FIRE PROTECTION PANEL JHilLP841 Activity Evaluated: ECN 29035; USAR Change 8-225                               Log Number: 98-151 Engineering Change Notice (ECN) 29035 adds a cutout to the front door of Panel IH13-P841 to facilitate access to the " Acknowledge & Silence" and " Reset" pushbuttons. The panel displays and annunciates fire alarms and trouble alarms, transmits this information to a printer, and
     . provides information to Control Room Operators about fire-related conditions within the plant's fire zones. This equipment is considered to be instrumental in the mitigation of the affects of the design basis fire. The modification to the panel door does not affect the operation or function of the panel, it only facilitates access to the operating pushbuttons. ' This modification does not affect compliance with National Fire Protection Association (NFPA) Code 72D for a central supervising station.; It only affects the Underwriter's Laboratories (UL) rating of the panel in that the panel with the cutout is not the tested configuration. This affects bounding statements regarding UL
: listing given in USAR Section 9.5.1, Appendix E, and NFPA Code Conformance Evaluation Checklist for NFPA 72D, Section 2022. However, no other aspect of the UL rating is affected.
This activity does not alter any assumption regarding the " postulated fire" for the fire area
: affected by this change as described in the Fire Hazards Anaiysis (Updated Safety Analysis Report (USAR) Appendix E). Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. In the event a malfunction of fire protection
    ' equipment monitorad by this panel occurs, the ability of the panel to exhibit a trouble alarm or detect a fire, and the authorized operator's ability to acknowledge the condition has not been               ;
altered. This modification does not reflect any change to the radiological consequences or safe I
i

T                w                                                                                                                      i
  .                                                                                                      Anachment 2                    l to U-603276                    i Page 159 of 252 shutdown actions associated with a postulated fire involving the Fire Protection System and other plant hardware, systems, or procedures. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This change to the panel only allows authorized access to the operating pushbuttons and
                       'does not alter the functional capabilities of the fire alarm system or the panel, affect UL requirements, or affect NFPA requirements against unauthorized use. Therefore, this change does not create the possibility of an accident or malfunction of equipment imponant to safety of a different type. All acceptance values or design limitations to the Clinton Power Station Fire Protection System have been previously deleted from the Technical Specifications and reflected in USAR Section 9.5.1, USAR Appendix E, and US AR Appendix F. Therefore, this activity does
                              ~
not reduce a margin of safety as defined in the basis for any Technical Specification.-                    -
CIRCULATING WATER PUMP BUILDING (KELLY MINI-SPAN ENCLOSURE)
                       . Activity Evaluated: ECN 31020; USAR' Change 8-220 '                            Log Number: 98-152-Engineering Change Notice (ECN) 31020 converts the existing building that covers the Circulating Water (CW) Pumps (installed by Temporary Modification 97-096, refer to Safety Evalua't ion 97-232) from temporary to permanent. The temporary power installed by Tempolary Modification 97-096, including security lighting, re' mains in place. The building is adequate for year-round operation of the CW pumps in all plant modes. The most likely path by which this change could initir.te an accident would be by collapsing the CW Pump Building under a wind or snow load in t-     causing the CW pumps to trip causing a plant trip. Expansion anchors, tension L tie bridging roos and channel connections are installed to ensure the structural adequacy of the CW Pump Building under snow, wind, and seismic loading conditions. The CW Pump Building is not designed for the Design Basis Tornado (DBT) and is not required to be. Its impact as a potential missile during a DBT is enveloped by the Tornado Missile Hazard Analysis, Therefore, this change does 'not increase the probability or consequences of an accident previously evaluated
                        'in the Updated Safety Analysis Report (USAR). The CW pumps located in the CW Pump Building are classified non-safety related and are not considered equipment important to safety.
                       ~ Postulated failures _ of this building through collapse or destruction by tornado are not postulated to damage neighboring structures containing the shutdown service water system which is safety related. Also, the overall system performance and system functionality are not affected by this
                        ' change, Fire protection features for the CW pump building provide a building egress path, do not            ;
                       > impact access to the Circalating Water Screen House (CWSH), and do not add flammable or combustible materials. Therefore, this change does not increase the probability or consequences of a malfunction of equipment imponant to safety previously evaluated in the USAR.~ This change makes Temporary Modification 97-096 permanent. Expansion anchors, tension tie bridging rods and channel connections are installed; temporary wire / cable bracing arrangement is removed; fire extinguishers are added; and details for building panel removal and restoration to enhance ventilation to maintain ambient temperatures are provided. None of these modifications
: change the functional or operational characteristics of the CW pumps or any other plant systems, structures, or components. Therefore, this change does not create an accident or malfunction of equipment important to safety of a different type. The Technical Specifications do not discuss the I

c                           >                       ._ -.                                                                                                ~ ~l
                                                                                        ''j-
                                                            ~
y l.        t                                                                                                            _
Attachment 2 2>                             '
4  ,
                              '                                                                                                to U-603276
', 'J Page 160 of 252
                          ,         a j                         , CW pumps or the CWSH. Therefore, this activity does not reduce a margin of safety as defitted 4                     .,        ' 'm the basis for any Technical Specification.            "
l                                      ; ['y                  .                                _    ,
f,                          m YEAR 2000 REACTOR RECIRCULATION FLOW CONTROL VALVE PLC UPGRADE                                                          l
     <.             4             '
L ACCEPTANCE TEST:
Y                  ,,                              ,.                                                                                           -
4I                                 "
                                               " Activity EvaluatSl:iCPS Procedure 2805.00                                     Log Number: 98-153
                                          +
p(
            ,            c M' 9' ' S two1Ger,eral Electric Programmable Logic Controllers (PLCs) are used to operate the Reactor Recirculation (RR). Hydraulic Power Units (HPUs). The PLCs provide input / output, control i                             1 functions land protective control operation of the RR Flow Control Valves (FCVs). This activity ilavolves the post installation acceptance testing of Engineering Change Notices 31129 and 31130
: (Reference Safety Evaluation 98-146) which upgrade the RR PLC Central Processing Unit (CPU) firmware from revision level 3.52 to '8.20. This test verifies that the new revision of the CPU
      .-                                    L firmware operates the original logic program meeting the functional requirements of the RR PLC j ,1                           . and supports the year 2000 compliance effort at Clinton Power Station. The proposed activity
                                               .does not affect the system performance but verifies that the upgraded PLC CPU operates
                                             - correctly,' meeting all alarm conditions, annunciations, and logic operations. This activity does not change. system response characteristics, cause system operation outside ofits design limits, cause operational transients in the system, or cause adverse system interaction with other systems.
Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The only events that are initiated by the FCVs are Recirculation Flow Control Failure - Decreasing Flow and Recirculation Flow e                  L Control Failure'- Increasing Flow while in Mode 1 or 2 operation of the plant. During the special
           ,          ,4                  L tet, no RR flow takes place and the plant is restricted to Modes 4 or 5. This limits the affect on the RR sydem with the associated FCV hydraulically isolated from its HPU. In addition, this
       ,,                                     ^ testing verifies correct operation of the PLCs in preparation for their operation in Modes 1,2, and 7                                        : 3. Therefore', this change does not increase the probability or consequences of a malfunction of
,                                           4 equipment important to safety previously evaluated in the USAR. The only component that has been altered in this activity are the PLC CPUs to correct a day of the week error that would occur L                                           , in the year 2000. The replacement PLC CPU firmware operates in the same manner as the old                      ,
revisioit / uter the HPU and FCV are restored to normal and alljumpers are verified to be
                                             ' removed with the system in a normal lineup, the only failures that could occur have been 4
previously evaluated. Therefore, this change do'es not create the possibility of an accident or
: malfunction of equipment important to safety of a different type than previously evaluated.
o                      LTechnical Specifications Bases 3.4.2 discusses the stroke rate of the valves based upon the
                                            ? hydraulic limitation where the limits are s 30% per second in the open direction and s 60% in the shut' direction. The PLC limits the change in flow rate to 11% per second and will fail the FCV Y                                    . "as-is" if this rate of change is exceeded assuming that the ladder logic program code functions as de's igned. Therefore, this activity does not reduce s margin of safety as defined in the basis for           !
any Technical Specification.

9
.                                                                                        Attachment 2 to U-603276            i Page 161 of 252 RHR HEAT EXCHANGER IB SERVICE WATER INLET FLOW LOOP RANGE RE-SPAN                                      1 l
Activity Evaluated: ECN 31096; USAR Change 8-229                                Log Number: 98-158 Engineering Change Notice (ECN) 31096 increases the Residual Heat Removal (RHR) Heat Exchanger (HX) IB Service Water'(SX) Inlet Flow Loop range of measurement from 8,000 gallons per minute (gpm) to 10,000 gpm. This flow loop measures the total SX flow through the following equipment: Emergency Core Cooling System RHR HX Room IB Cooling Coil Cabinet, Main Steam Isolation Valve Outb_oard Room Cooling Coil Cabinet, and RHR Heat                  1
       - Exchanger IB. Postulated accidents related to the RHR HX SX flow indication evaluated in               )
Chapters 6 or 15 of the Updated Safety Analysis Report (USAR) are not affected by this ECN.            I The only difference between the original flow transmitter and the new one, is the maximum calibration range. The new instrument indication range still meets the specified loop accuracy of 320 gpm for normal flow of 5800 gpm. The affected instmments are used only for monitoring             )
       . purpose and cannot initiate or automatically actuate any system or component. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in       i the USAR. _ Since the change only affects the instrument loop monitoring range, the RHR HX or         )
any other safety related equipment could not possible fail or degrade as a result of this change. A   j
       - loss ofindication would not prevent the automatic actuation and proper operation of the RHR system. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR; The new flow transmitter           .
i meets all approved design, material, and construction standards This plant change enhances the
        - RHR HX SX Inlet Flow monitoring capability for Ma in Control Room operators by prov ii         d ng a
        . wider monitoring range without affecting the required operating instrument accuracy. This plant change does not introduce any new type ofinstmment or equipment or modify existing instrument or equipment that might affect the function or mode of operation of the RHR system or the RHR HX SX Inlet Flow. Therefore, this change does not create an accident or malfunction of equipment important to safety of a different type than previously evaluated. Technical Specification Basis B3.6.2.3 states that during a design basis accident, a minimum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below the design limits. The SX flow assumed in the analyses is 5800 gpm per pump to the RHR heat exchanger. In order to' prevent damage to the RHR heat exchangers, the tube side flow should not be allowed to exceed 9000 gpm. This plant change increases the RHR HX 1B SX Inlet Flow Loop measurement range from 8000 gpm to 10,000 gpm. This enhances the RHR HX SX Inlet Flow monitoring capability for Main Control Room operators by allowing
        . the operator to determine proximity to maximum flow conditions. In addition, this change does not degrade the ability to determine flow rates within the existing indication band. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification. l l
v f

Attachment 2 to U-603276 Page 162 of 252
: REVISE THE DEFINITION OF RESTRICTED AREA AND RADIOLOGICAL CONTROLLED AREA Activity Evaluated. USAR Change 8-230                                            Log Number: 98-159 Updated Safety Analysis Report (USAR) Change Package 8-230 revises the definition of                             -
Restricted Area and Radiological Controlled Area _(RCA) as listed in USAR Appendix A to ensure fidelity with 10 CFR 20, Offsite Dose Calculation Manual, Sargent & Lundy Design Criteria, Technical Specification 5.5.4, and other USAR sections. This is an administrative
          - change to establish consistency between Clinton Power Station license basis documents, design requirements, and federal regulations. This change does not involve the operation, maintenance,
          - or design of equipment important to safety or its environment.' Also, this change does not impact any manual or automatic actions that would be taken in the event of an accident. This change has no impact on the Exclusion Area or Low Population Zone Boundaries nor does it have any impact on the radiation dose limits in the event of a design basis accident as given in 10 CFR 100.
Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This activity
          . does not introduce any new component or system into the plant, nor does it change any component or system already in the plant. In addition, this activity does not involve any physical changes to any boundaries. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated.
Technical Specification 5.5.4 provides the administrative and program controls to ensure compliance with Federal Regulations regarding the radioactive effluent dose limitations. This
          ' change is consistent with those requirements. This activity does not impact any acceptance values, design limitations, safety limits, or Limiting Conditions for Operation specified in the Technical Specifications. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
UPDATE ROOM USAGE FOR MCR HABITABILITY ENVELOPE Activity Evaluated: USAR Change 8-231                                            Log Number: 98-161 This change updates the room usage for the Main Control Room (MCR) Habitability Envelope in the Updated Safety Analysis Report (USAR). This change alters room assignments and nomenclature for occupants and contents. While room designation associated specifically with emergency supplies storage is being revised, required supplies and capabilities remain unchanged           -
and within the habitability envelope. No changes affect the "At the Controls" area of the MCR.
This activity i_ntroduces no credible failure modes that can act as an initiating event for a design basis event described in the USAR. This change does not reduce stmetural integrity, ventilation,
           ' electrical reliability, or personnel equipment required to be available in the event of an emergency.
These changes affect only how administrative activities are conducted during normal operations and do not involve the introduction of new materials. Therefore, this change does not increase the probability e ' n accident or malfunction of equipment important to safety previously evaluated in the : A The MCR Habitability Envelope is designed to minimize radiological

           $l 4 -                                                                                            Attachment 2 to U-603276 Page 163 of 252 exposure to less than 5 Rem whole body for the duration of an accident and less than 15 mR/hr average over 30 days. lIt also contains the facility appurtenances to suppon the operators for a
               ? duration of 30 days of abnormal operation and contains provisions for 7 operators for 5 days
_ without replenishment. The specific changes to room usage for these areas within the MCR Habitability Envelope does not compromise or in any way affect the ability of the MCR '
Habitability Envelope to ' perform its intended function. No envelope barriers are breached and the       .
ventilation system is unaltered.' Therefore, this change does not increase the consequences of an accident or malfunction'of equipment important to safety previously evaluated in the USAR. No
               ' changes to the ventt. tion system, electrical power reliability, envelope structures, or fire loading are being made as ; tuult of this change. i No new toxic materials have been introduced.
Therefore, this change does not create the possibility of an accident or malfunction of equipment
               ' imponant to safety of a different type. The Technical Specifications address the MCR ventilation             .
system and an air conditioning system. Ventilation system and air conditioning system performance is not affected by this proposed change; therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
CHANGE ACTUATOR MOTOR FOR SX ISOLATION VALVE 1SX014A TO INCREASE TORQUE THRUST
                 ' Activity Evaluatedt ECN 31173                                                   Log Number: 98-162 Engineering Change Notice (ECN)b 1173 replaces the motor in the actuator for Motor Operated Valve (MOV) ISX014A, Shutdown Service Water System (SX) isolation valve. The new motor is of the same design as the existing motor; except that it: develops 5 fl. Ib torque, rates at 0.33
                - horsepower, has an EB48 frame, and utilizes Type RH insulation. The Shutdown Service Water (SX) system provides a reliable source of cooling water for several safety related components wMen are essential for a safe shutdown of the reactor following a Design Basis Accident (DBA).
The Mer torque provided by the new actuator motor meets the design basis safety-related
                                                     ~
function toroue/thmst at system Degraded Voltage (DV) in accordance with Limitorque Technical Update (TU) 98-01.
l                    The SX system is'an accident mitigation system. The Updated Safety Analysis Repon (USAR) l                ' does not evaluate any accidents that are initiated by SX. 'This design change ensuwhe MOV
                                                                                                                             ^
l
                 ' actuator capability meet the safety related functional requirements. Therefore, timhange does L
                ~ not increase the probability or consequences of an accident previously evaluated in the US AR.
        >            Consistent 'with the existing actuator configuration, the larger motor will not over torque the valve int-rnals when the torque switch settings are within the requirements. Also, the larger motor has no adverse affect upon the Motor Control Center bus loading or diesel generator loading and is seismically acceptable. The larger motor requires more current during opening and closing the valve, calculations in the ECN confirm that there is no adverse impact on the electrical system by the addition of the larger motor. Therefore, this change does not increas . n.e
                 - probability of a malfunction of equipment important to safety evaluated previously in the USAR.
These design changes improve the performance of the valve / actuator assembly for postulated conditions and do not degrade the performance of the valve and its capability to provide isolation.
4
  \
                                                                                                                              )
I L                                                                                                                            1

                                                                                                                    )
                  't.
Attachment 2 to U-603276
                                      ,                                                Page 164 of 252
       ' Therefore, this change does not increase th5 consequences of a malfunction of equipment -
important to safety previously evaluated in the USAIL This design change replaces an actuator L motor with another motor of the same design and type, there are no new components being introduced into the plant. The failure modes are no different than the failure modes for the
     ,   existing motor; The MOV has been determined by calculation to operate ar.d achieve the thrust
       , requirements necessary to open and close the valve under both normal operating and accident              .
conditionssTherefore, this change does not create the ;% /Hiity of an accident or malfunction of
       . equipment important to safety of a different type. Tha sage m Stains the valve functional requirements by upgrading the motor. The configumm W 3h pd, however, there no changes to the acceptance limits, design basis, ftmetion or methou armcintaining acceptable flow rates.
Therefore, this acdvity does not reduce a margin of safety as defraed in the basis for any Technical Specification.
3- t     RHR HEAT EXCHANGER 1 A SERVICE WATER INLET FLOW LOOP RANGE RE-SPAN Activity Evaluated: ECN 31095                                                 Log Number: 98-163 i
Engineering Change Notice (ECN) 31095 increases the Residual Heat Removal'(RHR) Heat                       )
Exchanger (HX) 1 A Service Water (SX) Inlet Flow Loop range of measurement from 8,000 gallons per minute (gpm) to 10,000 gpm. This flow loop measures the total SX flow through the following equipment: Emergency Core Cooling System RHR HX Room 1 A Cooling Coil Cabinet,- Main Steam Isolation Valve Outboard Room Cooling Coil Cabinet, and RHR Heat
        . Exchanger I A. Postulated accidents related to the RHR HX SX flow indication evaluated in Chapters 6 or 15 of the Updated Safety Analysis Report (USAR) are not affected by this ECN.
       . The only difference between the original flow transmitter and the new one, is the maximum calibration range. The new instrument indication range still meets the specified loop accuracy of 320 gpm for normal flow of 5800 gpm. The affected instruments are used only for monitoring purpose and cannot initiate or automatically actuate any system or component. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in tbv NARJ Since the change only affects the instrument loop monitoring range, the RHR HX or y      aer safety related equipment could not possible fail or degrade as a result of this change. A le findication would not prevN the automatic actuation' and proper operation of the RHR system. :Therefore, this change does not increase the probability or consequences of a malfunction of equipment imponant to safety previously evaluated in the USAR. The new flow transmitter meets all approved design, material, and construction standards. This plant change enhances the RHR HX SX Inlet Flow monitoring capability for Main Control Room operators by providing a wider monitoring range without affecting the required operating instmment accuracy. This plant
: change does not introduce any new type ofinstrument or equipment or modify existing instrument or equipment that might affect 'the function or mode of operation of the RHR system or the RHR HX SX Inlet Flow. Therefore, this ch.mge does not create an accident or malfunction of equipment imponant to safety of a different type than previously evaluated. Technical Specification Basis B3.6.2.3 states that during a design basis accident, a minimum of one RHR             1 suppression pool cooling subsystem is required to maintain the primary containment peak pressure          j and temperature below the design limits.' The SX flow assumed in the analyses is 5800 gpm per 1

o 3                                                                                         Attac!unent 2 to U-603276 Page 165 of 252 7        ,
l               pump to the RHR heat exchanger. In order to prevent damage to the RHR heat exchangers, the
              . tube side flow should not be allowed to exceed 9000 gpm. This plant change increases the RHR -
HX 1B SX Inlet Flow Loop range of measurement range from 8000 gpm to 10,000 gpm. This p             : enhances the RHR HX SX, Inlet Flow monitoring capability for Main Control Room operators by L              . allowing the operator to determine proximity to maximum flow conditions. In addition, this l-            ' change does not degrade the ability to determine flow rates within the existing indication band.
Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical ~
              ; Specification.
;               SAR CHANGE TO MATCH PLANT CONFIGURATION FOR STANDBY LIQUID CONTROL DRAIN PATHS l
l              - Activity Evaluated: USAR Change 8-232                                         Log Number: 98-164 This change revises Updated Safety Analysis Report (USAR) Section 9.3.5.2 to reflect the change l               in the way the Standby Liquid Control (SLC) waste water is being routed. The USAR is revised L             Lt o statei " Equipment _ drains and tank overflow may be routed to separate drums that can be removed and disposed ofindependently to prevent any trace of boron from inadvertently reaching
              . the reactor. Alternatively, this drainage and overflow may be routed to the radwaste system where boron is removed and clean effluent returned , the cycled condensate system."
Previously, the USAR indicated that the drains and tank overflow could only go to the drums.
              .- The accidents evaluated in the USAR involve rapid increases in power from positive reactivity addition transients? The most probable failure associated with this change would involve the
              - introduction of boron into the cycled condensate returned from the radwaste system. This has been evaluated and determined to'not result in a greater than 50 parts per million (ppm)
               ; concentration of boron in the returned cycled condensate. Introduction of cycled condensate > 50 ppm of boron in the water would result in negative reactivity being added to the core and a '
                ~ decrease in power. . Recycled water boron concentrations will have little affect on reactivity since the possible concentrations are < 50 ppm. - Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The reactor is the equipment important to safety affected by borated water. Drainage or leakage from the SLC tank l
is controlled procedurally so that the water is processed via the floor drrN system. Boron is removed from liquid waste through evaporation, Liquid Radwaste Demineralizers, and Condensate Polishers. In the event there was an inadvertent addition of boron into the reactor water, the result would be a reduction in reactor power. Recycled water boron concentrations would have little affect on reactivity since the'possible concentrations are < 50 ppm. Therefore, E
this change does not increase the probability or consequences of a malfunction of equipment .
important to safety previously evaluatJ in the USAR. Since there are no uncontrolled sources of boron'that could contaminate the recycled water supply, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type.
L                Technical Specification Sections 3.1.7 and B 3.1.7 provide requirements for SLC system
               ; operability. The method of removing SLC' waste water has no affect on the SLC systen.
L                requirements discussed in these bases. Neither, the Technical Specifications nor the bases discuss

Attachment 2
                                                                                                         ~
to U-603276 Page 166 of 252         .
l the Liquid Radwaste system. Therefore, this activity does not reduce a margin of safety as defmed in the basis for any Technical Specification.
SHUTDOWN SERVICE WATER BOUNDARY VALVE LEAK TESTING Activity Evaluated: CPS Procedure 2804.00 R/0                                    Log Number: 98-168 This activity drains Shutdown Service Water (SX) system piping at various locations to determine the seat leakage across 12 SX boundary valves. The testing supports the operability of the Ultimate Heat Sink (UHS) by verifying that valves do not leak excessively; thereby, maintaining and conserving the volume of water contained in the UHS. During the testing, one division of SX is out of service; however, the other division is available for any cooling requirements. The SX system is considered an accident mitigation system. There are no accidents / events evaluated in Chapters 6 or 15 in the Updated Safety Analysis Report (USAR) initiated by the SX system.
Therefore, this change does not increase the probability or consecf sces of an accident previously evaluated in the USAR. The SX system provides a reliable source vt' cooling water for station auxiliaries which are essential for a safe reactor shutdown. During the performance of this activity, the applicable division of SX system is shutdown / unavailable / isolated and drained. The applicable drain valve is then opened to determine the applicable leakage past the valve being tested. If, during the test on valves ISX011 A/B, it is determined that leakage past either one of these valves is excessive, then applicable actions would be taken to preclude making the applicable available SX system unavailable. For the other valves, the other division of SX would be available for cooling. Therefore, this change does not increase the probability of a malfunction of equipment important to safety previously evaluated in the USAR. USAR Section 9.2.1.2 discusses two potential paths for radioactive inleakage to the SX system, the Residual Heat Removal (RHR) Heat Exchangers and the Fuel Pool Cooling and Cleanup (FPCC) Heat Exchangers. During the performance of this activity on the component cooling valves, the heat exchanger and associated piping is drained per applicable approved procedures. If a tube leak were to exist in the FPCC heat exchanger, it would have been identified and applicable corrective actions taken prior to performance of this activity. This activity will be initiated only afler verification that the redundant SX train is completely operable. No activities are involved in this testing that would result in a postulated failure or malfunction of the redundant SX train.
Therefore, this activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This activity does not introduce any new components into the plant. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type. Technical Specification 3.7.1 discusses " Division 1 and 2 Shutdown Service Water (SX) Subsystems and Ultimate Heat Sink (UHS)." Since this activity is only performed on one division of SX at a time and the other division is available for any cooling requirements; there are no design limitations affected by this activity. Further, this activity verifies that assumed performance standards for the SX system are met. Therefore, this activity does not reduce a margin of safety as definea in the basis for any           )
Technical Specification.

    .                                                                                   Attachment 2 to U-603276 Page 167 of 252 INSTALLATION OF HANDSWITCH FOR THE CONDENSER VACUUM PUMP DISCHARGE VALVE CIRCUIT B Activity Evaluated: ECN 30916                                                   Log Number: 98-172 This activity installs a two position spring-return switch in the condenser vacuum pump circuit to      -
allow the condenser separator tank discharge valve to be manually opened prior to starting the condenser vacuum pump. The discharge valve is controlled via an air-operated solenoid and fails closed on loss of air or loss of power. The Condenser Vacuum system is a non-safety related system powered from a non-safety related power source and is not credited for mitigating any evaluated accident. This activity enables the condenser separator tank discharge valve to be opened before the condenser vacuum pump is started. This sequence prevents expelling the loop             ,
seal of the tank. The only postulated time that a transient could be initiated as a result of this change would be when steam is being admitted to the condenser, a vacuum is present, and the
        . vacuum pumps are being prepared for initial start. The transient would be limited to a loss of condenser vacuum, but a check valve located upstream of the pump is provided to specifically prevent inleakage and prevents an actual vacuum loss. The spring-return switch is installed locally in a junction box with all wiring taking place inside the box. This change improves overall system performance since the loop seal is no longer expelled and no adverse affects on this system or any other systems are created. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). The two credible failure modes for the switch is failure in the contact open position or failure in the contact closed position. Failure of the spring-l         return handswitch in the contact open position would result in the discharge valve not opening, a conservative state. Failure of the spring-return handswitch in the contact closed position would l
!         result in the discharge valve remaining open, a non-conservative state. Failure of the valve in the l         open position will not initiate any sequence of events resulting in any type of accident not already l         bounded by the loss of condenser vacuum event. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than l         previously evaluated. There are no Technical Specifications associated with the Condenser Vacuum system including logic and controls, nor is this system or its performance related to any margin of safety. Therefore, this activity does not reduce a margin of safety as defined in the basis j          for any Technical Specification.
INSTRUMENT SETPOINT AND UNCERTAINTY CR OPERABILITY DETERMINATIONS AND EVALUATION Activity Evaluated: 1 98-07-208-OD; l-98-07-302-OD;                           Log Number: 98-173 1-98-07-377-OD; l-98-07-308-OD; l-98-07-308-OE-1 Operability Determination (OD) 1-98-07-308-OD and Operability Evaluation (OE) 1-98-07-308-OE-1 were prepared for Level I Condition Report (CR) 1-98-07-308, which documents an adverse trend ofinstrument setpoint calculations not performed in a controlled or

1
_,                                                                                              9 Attachment 2 to U-603276 Page 168 of 252 quality manner. The findings of the 1 98-07-308-OE-1 bound the conclusions of the 1-98-07-208-OD,1-98-07-302-OD, and 1-98-07-377-OD; and therefore, this 10CFR50.59                        )
evaluation is also applicable to the 1-98-07-208-OD,1-98-07-302-OD, and 1-98-07-377-OD.                  I The OD and OE were prepared to address operability ofinstmmentation functions required for
     . operation in Mode 4. The 68 parameters evaluated by 1-98-07-308-OE-1 included 16 instrument setpoint functions and 52 instrument indication functions and all 68 functions are considered important to safety. Ths setpoints have two credible failure modest setting / indication less
     ~ conservative than design and setting / indication more conservative than design. The indication functions by defini tion are not accident initiators, as they are used for operator indications and
                                                ~
: surveillance procedure acceptance devices. The OE re~sults document the indication functions are adjusted such that no instrument indications exceed the applicable allowable values defined by the applicable Mode 4 Technical Specifications. . Similarly, instrument setpoints r3 justed less conservatively than design are not considered to be accident initiatorsi An abnormal event or accident.would be required to establish plant conditions resulting in actuation of an instrument adjusted to a non-conservative setpoint value. The instrument setpoints and indications applicable in Mode 4 were conservatively established, and thus will not change, degrade or prevent actions assumed in the accident analysis. The instrument setpoints providing automatic equipment .-
     ~ initiation functions were determined to be adequately established for plant operation by the OE.
The original setpoint determinations have not resulted in a history of spurious equipment actuations, and the errors found by the OE did not result in any setpoints exceeding allowable
     ; values or approaching analytical limits. . Therefore, the setpoints validated or established by this .
     . activity will not change, degrade or prevent actions ' assumed in the accident analysis. The existing
     ' plant surveillance procedures and calibration and testing requirements are not altered by this activity; therefore, this activity does not impose more severe or increased ' testing requirements'.
This activity does not degrade design basis performance by affecting any environmental, seismic or separation design bases or criteria of any applicable systems. In addition, this activity does not     I reduce system redundancy or independence. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety
                                                          ~
previously evaluated in the Updated Safety Analysis Repon (USAR). This activity does not                 {
        ' introduce any new structures, systems,~or components into the plant or any new system interactions. Therefore, this activity does not create the possibility of an accident or malfunction
     . of equipment important to safety of a different type than previously evaluated in the USAR. The margin of safety, as defined in the context of this evaluation, is considered to be the difference between the analytical limits and the design failure point for the related equipment, system or           j function. Although the OE/OD determined a significant number ofinconsistencies in the
     ; instrument setpoint program, no changes to the Technical Specifications were identified. Also,
      .the OE/OD did not determine any setpoints that exceeded an allowable value or approached an t
     ' analytical limit as defined by the Technical Specifications. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
1 ''
i

  .                                                                                   Attachment 2 to U-603276 Page 169 of 252 HPCS DESIGN AND PERFORMANCE FLOW TEST Activity Evaluated. CPS Procedure 2801.00 R/0                                   Log Number: 98-174 This activity conducts Clinton Power Station (CPS) Procedure No. 2801.00 R/0,"HPCS Design and Performance Flow Test," to verify that the High Pressure Core Spray (HPCS) system and              -
HPCS pump meet design and performance specifications. The testing is conducted while the facility is in Mode 4 or 5. While performing CPS No. 2801.00, two flow testing teclaiques are used: Suppression Pool to Suppression Pool and Reactor Core Isolation Cooling Storage Tank to RCIC Storage Tank. In the shutdown mode, a minimum of two Emergency Core Cooling System (ECCS) subsystems are required to be operable, which can be met by other operable ECCS subsystems. Since testing is performed with the facility shutdown, there are no design basis accidents other than the Failure of Residual Heat Removal (RHR) Shutdown Cooling that could be affected, or credible failure modes that can act as an initiating event for the accidents or transients as a result of this activity. Because of system independence and separation, there is no postulated failure mechanism that could induce the failure of the RHR Shutdown Cooling System.
Based on several Updated Safety Analysis Report (USAR) statements applicable to HPCS testing, the activity does not change system response characteristics, cause system operation outside ofits design limits, cause operational transients in the system, or cause adverse system interaction with other systems. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The only new features of this test are the throttling of valves 1E22-F023, or F010 and F011, and the closure of IE22-F036, which renders the HPCS system both inoperable and not available. Because a minimum of two other ECCS subsystems are required to be operable, this does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. Technical Specification 3.5.2, ECCS Shutdown, applies to Qis test. Provided at least two other ECCS subsystems are operable in Modes 4 and 5, HPCS may be declared inoperable and out of service for this test. Therefore, this activit) does not reduce a margin of safety as defmed in the basis for any Technical Specification.
EQUALIZATION LINES FOR SX INLET AND OUTLET VALVES TO RHR HEAT EXCliANGER LOWER SETPOINT FOR SX PIPING THERMAL RELIEF VALVE Activity Evaluated: ECN 31288; USAR Change 8-248                                Log Number: 98-175 Engineering Change Notice (ECN) 31288 removes the cap and installs a 3/4" equalization line from the bottom drain of the Shutdown Service Water (SX) system inlet and odlet isolation vahres (1E12F014A and 1E12F068A) to the 18" SX line on the Residual Heat Removal (RHR) heat exchanger side of the valvo. The equalization line for valve 1E12F014A includes a new l      drain valve as well. This allows water between the two wedges of the valve to communicate with the SX water between the isolation valves. This activity also lowers the setpoint for thermal relief valve ISX208A from 200 psig to 175 psig. This supports the operability of the valve and the design pressure of the pipe. The SX system is an accident mitigation system and cannot initiate the accidents for which it is required to be operable. The equalization line and new drain valve

p                                                                                                                           ]
AttEchment 2            '
to U-603276.
Page 170 of 252
: was installed and tested to the same criteria as the piping of the SX system. ECN 31288 does not change the operating requirements of the valves; thus, it does not change the performance of the
           . system. The equalization line has no moving parts Thus, the only credible failure directly associated with this change is a small bore line break. A break in one of the equalization lines can be isolated by closing both the 1E12F014A and IE12F068A. The lower setpoint is well'above.                     1 the operating pressure and any anticipated pressure spikes for the SX system. All applicable -             - f codes and standards have been met during implementation of this change. SX valves IE12F014A-
           . and IE12F068A prevent leakage of radioactive contaminants resulting from an RHR heat -
exchanger tube leak. The routing of the equalization lines from the valve body to the SX piping on the heat exchanger side of the valves minimizes the potential leakage of contaminated water to
           . the lake. With the heat exchanger isolated after a tube leak, pressure from the RHR system acts -
on the backside of a flex wedge disk, increasing the sealing capabilities of the valve. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of -
                                                                                              ~
equipment important to safety previous!y evaluated in the Updated Safety Analysis Report. Since the equalization line is seismically analyzed and built to the same design / construction codes and standards as the associated SX system piping, the equalization line is no more subject to failure than the remainder of the system. Therefore, failure of the equalization line merely represent another point for failure, and is boundea by the presently analyzed floofling in the RHR heat exchanger room and loss of an RHR loop. Therefore, this activity does not create the possibility
              .of an accident or malfunction of equipment imponant to safety of a different type than previously evaluated. Technical Specification 3.7.1 requires both trains _of the SX system be operable to ensure the effective operation of the RHR system in removing heat from the reactor. As -
previously discussed, the likelihood of the equalization line failure is the same as for the existing piping, and the reduction of the thermal relief valve lift setpoint increases the margin of safety for the piping. A pressure locking condition of the isolation valves would prevent these valves from
           ' meeting the requirement for Technical Specification Surveillance Requirement 3.7.1.2. The result of this change makes the operation of the valves more reliable by removing the potential for locking of these valves. Therefore, this activity does not reduce a margin of safety as defmed in the basis for any Technical Specification.
: 480 VAC DISTRIBUTION, SUBSTATION G, K, L "AS BUILT" LOAD DISCREPANCY
           - RESOLUTION Activity Evaluated: . US AR Change 8-239                                        Log Number; 98-176 This Updated Safety Analysis Report (USAR) change revises USAR Figure 8.3-1, " Electrical Lbading Diagram," and Drawing No. E02-1 AP03, Revision K listed in USAR Table 1.7-1. The figure and table do not reflect changes made in non-safety related bus G, K, and L during i >
construction in 1983 through 1985. Drawing Nos. E02-1 AP06, Revision D and E02-1 AP08, Revision E reflect the correct configuration and should be incorporated into the figure and table.
This is an editorial change and does not affect system design, function o, method of performing functionsEThe existing loads were analyzed in the Calculation for Auxiliary Power System Analysis and constructed accordingly prior to obtaining the original Operating License. This change modifies these USAR drawings to correctly reflect the system configuration and design e,
N l

   .-                                                                                     Attachment 2 to U-603276 Page 171 of 252 bases. These components and their evaluated loading are incapable ofinitiating any of the USAR evaluated accidents and de not function in any mitigatory capacity. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This activity does not introduce any non-evaluated electrical bus load changes to substations G, K, or L or any changes in the operator interface with the substation components. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type. This activity does not alter the operation of any system described in the Operational Requirements Manual, limit use of any equipment discussed in the bases of the Technical Specifications, and continues to establish the system requirements to assure validity of statements provided in the USAR.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
MONITORING THE 5 VOLT POWER SUPPLY IN THE 1H13-P652, RACS PANEL Activity Evaluated: Temporary Modification 98-093                                 Log Number: 98-177 Temporary Modification 98-093 connects a recorder to the output of the PS-15 Volt Direct Current power supply in the IH13-P652 panel for Division 2 of the Rod Action Control System (RACS). The RACS is a subsystem of the Rod Control and Information System (RC&lS). The recorder monitors the output of the power supply to check for transient spikes and pulses and is located near the panel in a location that is at least 6 feet from any safety related or seismicly qualified equipment. This activity was installed during rod venting activities where the Reactor Mode switch was in the " Refuel" position. According to Updated Safety Analysis Report (USAR) Section 7.7.2.2.1, the RC&lS has no function associated with any design basis accident.
The safety function that RC&lS performs is to prevent control rod withdrawal errors during control rod movement and provides rod block functions that serve to minimize the consequences of a control rod drop accident. USAR Section 15A.6.5.3 states that the control rod drop accident       ,
is only applicable when in Modes 1,2, or 3. This activity is only allowed in Modes 4 and 5, so the     ]
control rod drop accident is not affected by this change. Even though RC&lS is not required for         j plant shutdown, RC&lS does perform functions that are important to safety. The function and objective of the system is not affected by this activity, and other safety system inputs that provide  i logic or field contact signals to the RACS are electrically isolated. Radiological release limits associated with the accidents and transients are not exceeded and no fission product barriers are challenged due to this activity. . Components and stmetures providing a mitigatory function during Modes 4 and 5 are unaffected by this activity. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The expected and abnormal transients and accident events analyzed in Chapter 15 and Appendix A of Chapter 15 of the USAR, envelope the failure modes associated with this activity. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated.
Technical Specification 3.3.2.1 and 3.9.2 are the Technical Specifications associated with this activity. This actifty does not acect the capability of the RC&lS to respond as evaluated in the USAR and has no specified design limit, thus it does not reduce margins of safety associated with

1                                                                                                  ,
Attachment 2           ,
to U-603276 Page 172 of 252 cladding integrity. :This activity does not impact safety limits,' limiting safety system settings, cladding temperature, or maximum cladding oxidation.. The presence of the recorder will not be
    - allowed during the Modes of operation where the RC&IS is required to be operable. Therefore, lthis activity does not reduce a margin of safety as defined in the basis for any Technical Specification.                                                                                   .
LADDITION OF ASSISTANT VICE PRESIDENT POSITION
    , Activity Evaluated:' USAR Change 8-240 '                                          Log Number: 98-178
: This change establishes an Assistant Vice President position reporting to the Vice President / Chief J Nuclear Officer. The duties of this position include: assisting the Vice President with his duties,
     , with emphasis on plant recovery and outage completion activities. This position may participate on committees and in activities providing guidance to the site staff; and, in the absence of the Vice President, may assume responsibilities of the office. Since this is an administrative organizational
    - change which does not impact plant design or operation or modify any system, structure, or componentithis change does not increase t_he probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR).c This change does not introduce any new failure modes or event sequences.
Therefore, this change does not create the possibility of an accident or malfunction of equipment imponant to safety of a different type. Technical Specification 5.2.1.c requires that a specified -
corporate executive have corporate responsibility for overall plant nuclear safety. The addition of an Assistant Vice President does not change that responsibility, and therefore, does not reduce a margin of safety as defined in the basis for any Technical Specification.
REMOVAL OF CHARLES TODD INDUSTRIAL UNIFORM SERVICES FROM THE USAR
    ? Activity Evaluated:' USAR Change 8-237 '                                         - Log Number:. 98-181 This change removes reference to the Charles Todd Industrial. Uniform Services from the Updated
    . Safety Analysis Report (USAR). The Charles Todd Industrial Uniform Services is no longer in business,'and the ARAMARK Uniform Services has purchased the facility and its contents.
     ' According to ARAMARK personnel the facility has been closed, chemicals are no longer on site, the helicopter pad is no longer maintained, and no ARAMARK personnel are on the site. USAR Section 6.4 addresses the systems which maintain the habitabihty of the Main Control Room. The hazardous chemicals stored at Charles Todd Industrial Uniform Service were not discussed due to its separation distance from Clinton Power Station as well as the types and quantities of the material located at the facility. The presence of a hazardous chemical storage location and helicopter pad being located offsite does not degrade nor prevent any plant stmeture, system, or componcat important to safety from performing its assumed accident mitigating function.
Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This revision to the USAR does'not introduce any new structures, systems, or components or any new failure modes or event sequences. Therefore, this change does not create the possibility of an accident or I
g

  .                                                                                       Attachmcnt 2 to U-603276 l                                                                                         Page 173 of 252 I
malfunction of equipment important to safety of a different type. The Technical Specifications do not discuss the personnel employed at the Charles Todd Industrial Uniform Services facility, the helicopter pad formerly used by the owner, and the hazardous chemicals which had been stored at the facility, since they posed no risk to the Clinton Power Station site or personnel. Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.
NON-CONFORMING RPS INVERTER EPA CARD Activity Evaluated: 1-98-11-299-OD                                           Log Number: 98-182 R/2 Operability Determination (OD) 1-98-11-299 documents the analysis used to demonstrate                 ,
Operability for a non-conforming Electrical Protection Assembly (EPA) within the Reactor              l Protection System (RPS) inverters with compensatory actions that limit operation to Modes 4 and
: 5. The function of the RPS power source (inverter or transformer) and EPA card is to ensure power of the correct quality is provided for operation of the Scram Solenoid Pilot Valves (SSPVs). The anomaly identified in EPA operation is an extension of the trip times to values beyond the original design basis, with the potential for the longer voltage or frequency transients before the EPA trips the breaker to cause damage to the SSPVs. All other aspects of the EPA operation have been verified through testing to function per design.
The design basis accidents and transients that need to be addressed are Rod Withdrawal Error-Low Power and Control Rod Maloperation. The credible failure modes are under/over               )
voltage and over frequency transients causing damage to the SSPVs through " chattering" of the        i SSPV internals, overheating of the solenoid mounted on the SSPVs, damage to the high power optical isolators, or electrical shorts. Regardless of the cause of the damage, the result of the damage is limited to premature opening of the SSPV, failure of the SSPV to open, or slow              j opening of the SSPV. In all of these cases, the result is a' 1/2 scram occurring prematurely or not   i occurring as required. Since the scram is used to mitigate the accidents, a premature scram or failure of the scram to occur cannot be an accident initia*or. The primary equipment important to safety potentially affected by the non-conforming EPA are the SSPVs and related circuitry.
Analysis demonstrated that for the trip times for voltages from 85 VAC to 160 VAC with frequency variations f.. n 57 to 63 Hz, the SSPVs are not adversely affect.d. In addition, the SSPVs are capable on sithstanding a single low or high voltage transient without adversely affecting the ability for a 1/2 scram to occur; and thus, this activity does not degrade the performance of the SSPVs below the design basis. While the results of the non-conformance is the same for both EPAs, there is no link between the power source-EPA sets. Thus, a failure in one power source-EPA system does not result in a failure in the other power source-EPA system.
This activity will not change, degrade, or prevent actions described or assumed in the accident l
analysis for mitigating the effect of the transient. Thus, the activity cannot result in a change in the radiological consequences of the accidents and transients associated with the activity; and the assumptions of no radiological release and no fission product barriers being challenged remain intact. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis

e Attachment 3          ,
to U-603276
                                                                                               - Page 174 of 252 -
      -- s    Report (USAR). This activity does not degrade the performance of the SSPVs and related '
circuitry below the design basis; and, the functionality of the circuitry will be verified prior to
: using it again for SSPV operability via testing. This activity does not add any new structure, system, or component or any new system interfaces. Therefore, this activity does not create the
_ possibility of an accident or malfunction of equipment of a different type than previously evaluated in the USAR.' This activity does not affect any Technical Specification, safety limit,           -
limiting safety system setting, or limiting condition for operation (LCO) because this activity limits operation to Modes 4 or 5 where LCO applicability does not apply. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REASSIGNING QA ORGANIZATIONAL RESPONSIBILITIES ActivityEvaluated: USAR Change 8-224-                                             Log Number: 98-183
                      =
This Updated Safety Analysis Report (USAR) revision eliminates the Director - Quality Assurance (QA) position and adds the Supervisor - Quality' Assurance Operations and the '
Supervisor - Quality Assurance Maintenance positions on Figure 13.1-1. -The Supervisor positions are existing positions that are being shown on the figure as the result of the change.
Based on a separate 10CFR50.54a evaluation, it was concluded that this change does not reduce commitments 'or the effectiveness of QA functions specifically delineated in the QA Program Description.'. Due to the present staffing level of the QA department, the Director position has proven to be unnecessary. This change does not affect any systems, structures, or component
           - design, material, construction standard or method of operation. ' No changes to the facility are required due to this change.. Since this is an administrative organizational change which does not
           ~ impact plant design or operation, this change does not increase, nor create, the probability or consequences of an accident _or malfunction of equipment important to safety. Technical Specification'5.2.1 specifies the site organization be documented in the USAR. This change revises the reporting chain for the QA department; the US AR continues to contain the
            ' information required. Therefore, this activity does not reduce a margin of safety as defined in 11 e basis for any Tect.nical Specification.
RPS INVERTER EPA OVERVOLTAGE AND UNDERVOLTAGE SETPOINT CIMNGE
             ' Activity Evaluated:. ECN 31306 and ORM Change 23-7                                Log Number: 98-184 xThis activity involves changing the Overvoltage and Undervoltage setpoints to the Reactor Protectiori System (RPS) Inverter Electric Protection Assembly (EPA) circuit setpoints. The
            -- EPA voltage setpoints at the inverter are based on providing required voltage to the Main Steam
              ' Isolation Valve (MSIV) solenoids and the RPS scram solenoids when taking into account associated voltage drops.' The overvoltage setpoint is to protect the solenoids from overheating and/or damaging their insulation. The undervoltage setpoint is to protect the solenoids from chattering and potentially losing their control capability, resulting in a loss of primary scram action. Th'ese new setpoints ensure that the solenoids perform their intended safety functions.             j I
I 1
I

n U        ..
N I      E       ,                                                                                Attachment 2 L                                                                                               to U-603276 Page 175 of 252 Failure of the inverter could cause a loss of voltage output or voltage fluctuating outside ofits preset value. In either event, the inverter failure would be considered a single failure, With 'a
            " single failure of an invener, the EPAs are assumed to operate properly and would isolate the bus if the voltage values exceed the new setpoints. Regardless of the cause of the inverter failure, the result is limited.to premature opening of the A or B solenoids. The result is a half scram and
            ' de-energizing 1 of the 2 MSIV solenoids which is the same result as the previous licensing basis.      .
Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the Updated Safety Analysis Report (USAR), it is important to note, that the EPAs L
function on;y when'the inverter fails. It is not a credible scenario' to assume that both will fail.
The worst case scenario is that an invener fails causing a half scram and de-energizing 1 of the 2 -
MSIV solenoidsf The new setpoints cannot cause an inverter to faili Therefore, this change does          l increase the probability or consequences of a malfunction of equipment important to safety                l previously evaluated in the USAR. Since one inverter failure a the single failure, the other              l d
            - independent inverter and EPA would be available to perform its safety function when required.
:           . The new setpoints ensure that a trip would take place prior to causing damage to the scram L             solenoids and MSIV solenoids. Since this activity only involves an internal change on the EPA-cards end the new setpoints are adequate to protect the required loads, no new accidents or L;             equipment malfunctions are created. - Technical Specification 3.3.8 2 discusses the " Reactor -
             . Protection System Electric Power Monitoring System.". The overvoltage and undervoltage EPA trip setpoints are pan of the electric power monitoring systemi To maintain the margin of safety, the scram and MSIV isolation functions need to be maintained. Supplying the correct voltage to the downstream loads and isolating the bus if the voltage requirements are not met ensures the scram and MSIV closure function are not compromised. The most limiting voltage requirement and associated line losses determine the settings of the electric power monitoring instrument channels. The settings are calculated based on the loads on the buses and associated Uninterruptable Power Supply or alternate power supply being 120 VAC at 60 Hz. Taking into account the line losses, the new setpoints adequately protect the' solenoids from an overvoltage I
and undervoltage' condition by isolating the bus. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
L              REVISION TO THE REMOTE SHUTDOWN PROCEDURE L Activity Evaluated: CPS Procedure'4003.01                                         Log Number: 98-185
                                                                                                                       ]
This change revises Clinton Power Station (CPS) Procedure 4003.01, " Remote Shutdown (RS),"              ;
1t o include additional manual actions prior to evacuating the Main Control Room (MCR) in the event of a fire. These actions include initiating a scram in a different fashion, shutting the Main Steam Isolation Valves and Main Steam Line drains, terminating Feedwater injection, de-y                energizing the Division 1 and 2 safety relief valve solenoids, closing the Reactor Core Isolation L                Cooling (RCIC) Steam Supply valve, and closing the RCIC Turbine exhaust drain line valve.
None of these manual actions replace an automatic action rather they are performed to assure that any spurious actuation due to the postulated MCR fire does not result in the inappropriate operation ofequipment or' systems not required for Safe Shutdown. These actions are consistent l

Attachment 2            ,
to U-603276 Page 176 of 252 with the criteria identified in Nuclear Regulatory Commission Supplemental Safety Evaluation Report 6 and 10CFR50 Appendix R.
Revising CPS 4003.01 allows CPS to more accurately execute the Safe Shutdown requirements of 10CFR50 Appendix R. As defined in Updated Safety Analysis Report (USAR) Appendix F, Safe Shutdown Analysis (SSA), Section 1.2, the SSA demonstrates that for a fire in any single plant fire area, at least one method exists that is free of fire damage to achieve and maintain a safe shutdown condition. No fire hazards, combustibles, or ignition sources are added by this activity.
Therefore, this change does not increase the probability or consequence:, of an accident previously evaluated in the US AR. This activity clarifies and adds available manual actions to assure the timely operation, or preclude the inadvertent operation, of the equipment or systems involve due to a MCR fire. This activity does not involve a change to the design, function, operation, or testing of any equipment or systems affected by this change. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the US AR. This activity does not involve any new equipment or a change to the design, function, operation, or testing any existing equipment or systems affected. This activity does not reflect any change to or add any credible failure modes associated with any postulated fire involving the Fire Protection System and other plant hardware, systems, or procedures. Therefore, this change does not create the possibility of an accident or malfunction ofequipment important to safety of a different type. Technical Specification Section 3.3.3.2 discusses the Remote Shutdown System. There is no change to the RS system as documented in the Technical Specifications, and the RS system continues to provide the Control Room Operator with sufficient instrumentation and controls to place and maintain the plant in a safe shutdown condition from a location other than the Control Room. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specificition.
REVISION OF OFFSITE DOSE CALCULATION MANUAL Activity Evaluated: ODCM Revision 17                                            Log Number: 98-186 Revision 17 of the Offsite Dose Calculation Manual (ODCM) makes the following changes:
reformats various formulas in Sections 2 and 3 used for dose calculations to match the format of
  . the Efiluent Management System (EMS), change the calculational methodology in Section 9.0 to that used by the Meteorological Information and Dose Assessment System (MIDAS), and makes several editorial changes. The ODCM provides the methodologies and parameters used by Clinton Power Station to assure compliance with the radioactive efiluent dose limitations. The
,  applicable accidents evaluated in the Updated Safety Analysis Report (USAR) deal with dose to unrestricted areas from radioactive efiluents. The changes incorporated by this revision are not accident initiators. The editorial changes correct typographical errors, remove extraneous information, and clarify the intent of the ODCM. These changes do not alter or delete any commitment in the ODCM. The EMS software performs the actual dose calculations; chemistry personnel have compared the formulas used for EMS with.those used in Revision 16 of the ODCM. Verification and validation activities associated with EMS have shown that the EMS software provides doses equivalent to those received using Revision 16 methodology. The I

r                                                                                                            1 j
    .                                                                                 Attachment 2 to U-603276            i Page 177 of 252 MIDAS related changes incorporate changes in the methodology used to handle meteorological data used to calculate dispersion factors for cJfsite dose. The changes made do not alter commitments or capabilities related to efiluent control and offsite dose calculation. These changes do not impact the determination of radiation monitoring setpoints nor do the changes           ,
affect the operation of any other plant systems. Therefore, this activity does not increase the        )
probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The changes do not affect the operation of the Area
                                                                                                           .j a
I Radiation / Process Radiation Monitoring system since the setpoint methodology is not changed.
The change does not alter any monitoring requirements already in the ODCM. These changes are unrelated to any controls or operations of any plant equipment. Therefore, this activity does not       .
create the possibility of an accident or malfunction of equipment important to safety of a different   j type than previously evaluated in the USAR. The changes incorporated in this revision of the           {
ODCM are in full compliance with Technical Specifications for calculating doses to unrestricted        {
areas as a result ofliquid and gaseous effluents. The software programs have been tested and           j found to provide accurate results for offsite dose calculation. The new methodology introduced         j in Revision 17 is a different but eauivalent way of falfilling regulatory requirements for this         i activity. Therefore, this activity does not reduce a margin of safety as defmed in the basis for any   i Technical Specification.
ENVIRONMENTAL CONDITIONS IN AREAS SERVED BY CONTROL ROOM HVAC SYSTEM (VC)
Activity Evaluated: ECN 31313; USAR Change 8-260                                Log Number: 98-187 Engineering Change Notice (ECN) 31313 modifies the Clinton Power Station Updated Safety Analysis Report (USAR) by deleting the nominal ambient temperature and humidity conditions for         j the Control Room envelope and adding the limits of parameters which are required to be maintained for equipment operability. The Control Room envelope and its Heating, Ventilating and Air-Conditioning (HVAC) System (VC) are designed to assist in the mitigation of postulated accidents identified in the USAR. This activity changes the Control Room's environmental                l conditions arJ ts HVAC system operating parameters, with no direct or indirect impact on initiation of design basis accidents. The Control Room temperature is maintained by the VC system below the Technical Specification limit of 86 F. Minimum and maximum allowed relative humidities inside the Control Room are bebg revised from 35% to 5% for minimum and from 100% to 60% for maximum. The revised humidity levels are in accordance with the value specified in the equipment purchase specification. Revision of temperature, humidity, and control tolerance does not alter the design constraints for the control room and its HVAC. The Control Room envelope is sufliciently shielded and its habitability system are provided with components so that under accident conditions, exposure of Control Room operators does not exceed General Design Criteria 19 criteria. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. These changes will not degrade or prevent actions or impede access to vital areas for mitigating the afTect of an accident or transient.' This change clarifies the design capability of the VC system and the environmental conditions inside the Control Room envelope and the components therein. This change does not

I                                                                                      '
Attachment 2           ,
to U-603276
  -i:                             J                                                                   Page 178 of 252
: y. ,       alter the function or performance capability of any structures, systems, or components.
Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluatedc Technical Specification 3.7.4,
                      " Control Room Air Conditioning (AC) System," controls the~ operability of the VC system: This
                - . activity does not change these requirements, nor does it change the design or operation of the -
                   ' Control Room _ Air Conditioning' system. Therefore, this activity does not reduce a margin of             -
                   - safety as defined in the basis for any Technical Specification.
RESIDUAL HEAT REMOVAL (RHR) HEAT EXCHANGER SHELL SIDE TEMPERATURE
                   -RANGEc                >
c
                    ; Activity Evaluated:J 1-98-12-274-OE                                             Log Number: 99-001 This activity evaluates the. potential for the Residual Heat Removal (RHR) heat exchanger shell
                   . side (i.e.g RHR coolant side) temperature to drop below the design specification and Updated Safety Analysis Report (USAR) value of 40 F. This evaluation is necessitated by the very low
<                   : decay heat generation rate within the reactor and the potential inability to maintain the RHR Coolant Temperature greater than 40 F. The function of the RHR heat exchanger is to provide for unit cool down in 20 hours to a temperature at or below 125'F. Minimum temperature limits for the reactor are controlled by bypassing the RHR heat exchangers when the reactor
                     . temperature decreases to 70 F. System and operator responses to reactor water temperature decreasing to 70*F is not affected by a change in the minimum RHR heat exchanger shell side i temperature limit. This activity does not affect system performance or create adverse system interactions! Excessively cooled RHR water introduced into the reactor would result in a positive
                    ' reactivity insertion. If the operator did not act to control the power level, a high neutron flux
                      ' reactor scram would terminate the transient without violating fuel thermal limits and withou't any
                    - measurable increase in nuclear system pressure. These reactivity considerations are addressed
                     . separately. If the RHR heat exchanger malfunctions with no' appreciable decay heat in the reactor and the lake temperature near 32 , then the function of removing heat from the reactor is not needed, because there is no heat in the reactor. If the reactor is generating decay heat, then the RHR heat. exchanger inlet, outlet, and bypass valves can be operated per Clinton Power Station procedure 3312.03, "RHR - Shutdown Cooling (SDC) and Fuel Pool Cooling and Assist                         ,
            ,          (FPC&A)," to cool the reactor as needed. This activity does not affect the ability of the RHR
: pumps to move water from the suppression pool or reactor through the RHR system piping and
                    ' back to the~ suppression pool or reactor. Therefore, this activity does not increase the probability
                    .. or. consequences of an accident or malfunction of equipment important to safety previously                j evaluated in the USAR? The accident that could happen with this change is if a condition was created ir the RHR heat exchanger where the temperature sub-cools below 32 F and ice fctms in             ,
: the RHR heat exchangers and block the flow of water through the RHR heat exchanger shell side.              !
This is not credible given the ambient air temperature of the RHR rooms in the plant and the heat
                    - picked up by the piping. No new components or systems are introduced as a result of this change.
Therefore, this activity does not create the possibility of a accident or malfunction of equipment
                      .important to safety of a different type than previously evaluated in the USARL The Technical Specifications do not address the temperature range for the RHR heat exchangers. Technical

f:                                                                                                          \
l  .                                                                                   Attachment 2         <
!                                                                                     to U-603276           )
Page l79 of 252       l l    Specification 3.4.9 requires RHR' shutdown cooling to be operable under various plant conditions.
The change in RHR heat exchanger shell side minimum temperature supports the operability of RHR. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
ADDITION OF 8 HR BATTERY PACK LIGHTING TO SUPPORT MANUAL RESTART OF MCR HVAC CHILLERS l    Activity Evaluated: ECN 31208; USAR Change 8-244                                  Log Number: 99-002 This activity adds three battery pack light assemblies, associated lamps and conduits at elevation 825' of the control building. These battery packs support the manual restart of the Main Control Room (MCR) Heating, Ventilation, and Air Conditioning (HVAC) chillers and the manual operation of the makeup valves from the Shutdown Service Water system by providing lighting during a loss of offsite power. The battery packs are eight hour rated with eight corresponding lamps. The installation includes supports for the conduits, battery packs, and lamps. The power supply for these battery packs is supplied from Regular Lighting Cabinet (RLC) numbers 153, 154, and 155. During a loss of offsite power, the power to the RLCs is not available, and the l
battery pack lamps come on and illuminate the area. If the battery pack lights are connected to the standby lighting cabinets (SLCs), the battery pack lights are on only upon loss of power to the SLCs, which are , awered from the diesel generators. Since the area has minimal standby lighting fixtures, the lighting from these fixtures is considered insufficient to perform the required activities; and thus, the battery pack lights are required to be fed from the RLC in order to perform the intended function.
All components are similar to previously installed componants and will not create a new failure mode. The additional electrical load due to the battery pack:. is negligible and will not impact the ability of the Auxiliary Power System to perform its intended function. No accidents described in Updated Safety Analysis Report (US AR) Chapter 15 including " Loss of AC Power" are impacted by this design change. This change to provide " battery lighting" for the MCR chillers does not impact MCR habitability or design requirements. The potential fire protection program impact of the additional batteries has been evalua:ed and determined to be acceptable. Therefore, this
     - change does not increase the probability or consequences of an accident previously evaluated in the USAR. No rquipment or system, required for the safe operation and shutdown of the station is impacted by this design change. Therefore, tuis change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. There are two potential failure modes of the battery packs; the lights will not work or failurd of the supports. Failure ot'the lights will not affect system function, but could reduce l      lighting levels in the area to support manual function of the equipment. Failure of the battery      i pack supports could result in the battery pack falling during a seismic event; however, based on a field walkdown, the battery packs will not fall directly on safety related equipment. This design change adds equipment of a similar type as equipment installed and in use. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type. There are no Technical Specifications that discuss Appendix R battery pack L

Attachment 2            ,
to U-603276 Page 180 of 252 lighting. The battery pack lights do not impact the Auxiliary Power Systems or the diesel generator capabilities. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
DEGRADED VOLTAGE CONTROL CIRCUIT UPGRADE ON THE DIVISION III SX SYSTEM MOV BYPASS RELAYS Activity Evaluated: Modification AP-035 Supp 3,                          Log Number: 99-004 R/l USAR Change 8-147 Modification AP-035 Supplement 3 reconfigures the electrical circuit of the motor operated valve (MOV) overloa ' bypass relays (95-SX3 A, B & C) in order to stagger the pickup of these relays.
Originally, the relays were all connected in parallel and were simultaneously energized by the operation of a test switch. This modification reconfigures the 95-SX3B relay to be interlocked from a normally open contact of the 95-SX3 A relay. Relays 95-SX3A and 95-SX3C remain in their original configuration. This action improves available voltage to the relays, thus, increasing their reliability. This modification changes the control circuit to improve proper operation of the relays in the circuit, it does not change the function or method of operating the control circuit.
This modification does not change any systems that are design basis accident initiators in Chapter 6 or 15 of the Updated Safety Analysis Report (USAR), it only changes the control circuitry of certain accident mitigation or support systems. The reliability of the modified control circuitry willimprove the ability of the relays to perform their intended design function. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. This change introduces the addition of another contact to the control circuitry feeding one of the 95 relays, but the functioning as a unitized group has not changed. This modification improves the reliability of the relays in the modified control circuit to perform their intended design funct;en during normal voltage conditions. Periodic and post modification testing verifies the operation of the modified circuits. Components used in this modification meet the same design, material, testing, and quality standards as the original components. Therefore, this change a   does not increase the probability or consequences of a malfunction of equipment important to s safety previously evaluated in the USAR. The redesigned control circuit continues to perform its intended design function, so there is not an introduction of an accident of a different type than previously evaluated. This change improves the voltage to the relays, thereby improving their function to act as mitigators to a potential accident. No other component changes are involved.
Therefore, this change does not create the possibility of a malfunction of equipment important to safety.of a difTerent type than previously evaluated. During the installation and testing portions of this modification, the Unit is in Cold Shutdown and the affected equipment is de-energized at the breaker and tagged out, Applicable LCOs, procedures, installation specifications and post-modification testing is adhered to, to ensure the modification is installed in accordance with the design. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.

  .                                                                                     Attachment 2 to U-603276 Page 181 of 252 EQUALIZATION LINES FOR SX INLET AND OUTLET VALVES FOR RHR HEAT EXCHANGER, LOWER SETPOINT FOR SX PIPING THERMAL REU N VALVE Activity Evaluated: ECN 3033!                                                     Log Number: 99-005 This change does the following: 1) removes a cap from the bottom drain line of Shutdown Service Water (SX) inlet and outlet isolation valves, IE12-F014B and IE12-F068B, and installs a 3/4" equalization line from the bottom drain of the valves to the 18" SX line on the 7esidual Heat Removal (RHR) heat exchanger (1E12 B001B) side of the valves,2) includes a net drain valve for the equalization line for valve IE12-F014B, and 3) lowers the setpoint for reliu valve 1SX208B from 200 psig to 175 psig. These changes support the operability of these valves by preventing possible pressure locking and by limiting the pressure of the fluid within the pipe.
l The SX system is an accident mitigation system and does not initiate any accidents evaluated in           j the Updated Safety Analysis Report (USAE The equalization line and new drain valve were                  j installed and tested to the same criteria as ta piping of the SX system. This change does not alter       ;
the operating requirements of the valves, and therefore, does not change the performance of the          l system. Failure of the new piping is bound by the analysis for the 18 inch SX line. The                  )
installation of these lines do not adversely affect the flow requirements of cooling water for the RHR heat exchanger. The lower relief valve setpoint protects the SX piping from                           ;
overpressurization and is well above the system operating pretsure and any anticipated pressure           j i
spikes. The design function of the inlet and outlet isolation valves remain the same. The change ensures the operation of the inlet and outlet isolation valves by preventing a pressure locking condition. Therefore, this change does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the USAR. SX valves IE12-F014B and             i I
lE12-F068B prevent leakage of radioactive contaminants resulting from an RHR heat exchanger tube leak. The routing of the equalization lines from the valve body to the SX piping on the heat exchanger side of the valves preserves the barrier against potential leakage of contaminated water to the lake. With the heat exchanger isolated af er a tube leak, pressure from the RHR system acts on the backside of a flex wedge disk, increasing the scaling capabilities of the valve.             l Therefore, this change does not increase the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The equalization line has no moving parts; thus, the only credible failure directly associated with this change is a small bore line break Breakage of this line can impact the valves, the SX system piping, the RHR heat exchanger, and any equipment in the zone ofinfluence associated with the flooding and water spray. Failure of the equalization line merely represents another point for failure, and is bounded by the presently analyzed flooding in the RHR heat exchanger room and loss of an RHR loop.
Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type. Technical Specification 3.7.1 requires that two trains of the SX system be operable to ensure the effective operation of the PsHR system. The bases state a subsystem is considered operable when the associated piping, valves, instrumentation, and controls required to perform the safety function are operable. The likelihood of the equalization line failure is the same as for the existing piping, and the reduction of the thermal relief valve lif t

p                                                    -
Attachment 2           .
to U-603276 Page 182 of 252 setpoint increases the margin of safety for the piping. Therefore, this activity does not reduce a -
margin;of safety as defined in the Technical Specification.
DELETION OF MAKEUP CONDENSATE / CYCLED CONDENSATE HOSE STATION
       -VALVES
      . . Activity Evaluated: ECN 31315; USAR Change 8-324                              Log Number: 99-006 Engineering Change Notice (ECN) 31315 deletes the Makeup Condensate (MC) and Cycled Condensate (CY) hose station valves, OMC081G and OCY075, from drawings M05-1042, Sheet
      - 3, Revision Y and M05-1012, Sheet 5, Revision R. The change to M05-1012, Sheet 5 impacts Updated Safety Analysis Report (USAR) Figure 9.2-11, Sheet 5, which depicts the CY system.
      - These valves are no longer physically installed in the plant, instead, the piping has been capped.
The CY. system has no nuclear safety related function, except for the piping and valves form the containment isolation boundary. Failure of this system, with the exception of the containment
      - isolation valves (which this change does not affect), does not compromise any nuclear safety related system or component and 'does not prevent the safe shutdown of the reactor. The MC
      - system does not perform or assist in performing any nuclear safety function such as the safe shutdown of the reactor or to mitigate the consequences of a nuclear accident. Therefore, this change does not increase the probability or consequences of an accident or malfunction of-
      - equipment imponant to safety previously evaluated in the USAR. The piping at the hose station has been capped, thus preventing the possibility of flooding the Radwaste Building. A failure of this pipe cap would result in flooding, however, this type of failure has already been discussed in
      ; the USAR under the Internal Flooding section. . Therefore, this activity does not create the possibility of an accident or malfunction of equipment imponant to safety of a different type. The
      - Technical Specifications do not discuss the MC or CY systems; Therefore, this activity does not
      . reduce a margin of safety as defined in the basis for any Technical Specification.
CHANGE NORMAL LINEUP OF IA AND MC FLUSH ISOLATION VALVES FOR RADIATION MONITORS FROM OPEN TO CLOSE Activity Evaluated: ECN 31365; USAR Change 8 252 Log Number: 99-007 Engineering Change Notice (ECN) 31365 revises the valve lineup shown on Updated Safety Analysis Report (USAR) Figure 9.3-2 to show the Instrument Air (IA) and Makeup Condensate
      - (MC) flush isolation valves, for process and effluent monitors, normally closed rather than normally open. Changing the normal position of the IA purge supply or MC flush supply valves from'open to closed does not affect the operation, maintenance, or design of the process radiation or continuous air monitors.1 Nor does it degrade or prevent actions required for mitigating an
      > accident and does not affect or introduce new transients? The monitors provide continuous monitoring of radioactive effluent release paths and are not designed to mitigate the consequences of an accident. By changing the normal position from open to closed, it is less likely that effluent being monitored will become diluted, meas'uring less radiation than is actually present in effluent releases.' The change ensures the readouts from the monitors is more representative of the 4                                                                                                                f

  .                                                                                      Attachment 2
;                                                                                        to U-603276 Page 183 of 252         i efIluent. Therefore, this change does not increase the probability or consequences of an accident          !
or malfunction of equipment important to safety previously evaluated in the US AR, nor does it             l create an accident or equipment malfunction of a different type. The monitors affected by this             I change monitor efliuents for compliance with the Offsite Dose Calculation Manual and are not discussed in the Technical Specifications. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.                                          -
USE OF APPENDIX R PORTABLE EMERGENCY LIGHTING I
Activity Evaluated: ECN 31344; USAR Change 8-251                                    Log Number: 99-009 Engineering Change Notice (ECN) 31344 provides for portable, sealed-beam, 8-hour battery pack lights to be located in the corridor (elevation 801'-9") west of the Main Control Room (MCR)
      .and at the Remote Shutdown Panel (elevation 781')in the Auxiliary Building to' provide Control            j l
Room Operators with portable lighting in the event of a fire in the MCR. The portable lighting provides lighting for one-time manual actions associated with Safe Shutdown and the tripping of the High Pressure Core Spray pump. The lights are rechargeable battery type, with adjustable               i lamp to permit the light to be placed and adjusted allowing both hands to be free for the required         !
manual operations. A wall unit consists of four portable lights mounted on an aluminum plate and           l I
supported by seismically-rated unistruts. The wall units are equipped with chargers that are powered from the Regular Lighting Cabinet.
The portable emergency lighting is in compliance with 10CFR50 Appendix R. The addition of portable lights provides a minimal increase in the combustibles in the zones and has no affect on the fire classification of these zones, fire barriers, or fire-rated penetrations that prevent the fire from spreading to another plant area. Therefore, this change does not increase the probability or         l consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). This' activity does not affect the availability, performance, or reliability of safety systems, and there is no increased challenges to safety systems. The use of portable emergency lights are pan of the defense-in-depth of the design and programmatic fire protection features that ensure that the safe shutdown capability is maintained.
These features include the fire protection administrative controls, physical design of the Fire Protection System, manual fire suppression by the fire brigade, and emergency response procedures. The use of Appendix R-qualified portable lights enhances the defense-in-depth capability, and therefore, do not increase the probability or consequences of a malfunction of equipment imponant to safety previously evaluated in the USAR. A failure of a portable light used for a manual action to mitigate the afTects of multiple spurious actuations due to a fire could be compensated by the use of a spare ponable light. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a ditTerent type. The Technical Specifications do not discuss the Fire Protection Program. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.

                                                                                       . Attachment 2 L                                       '
to U-603276 Page 184 of 252 l-FIRE ALARM STROBE LIGHTS FOR PANELS 1H13-P661 & IH13-P662' Activity Evaluated: ECN 31341 & 31342                                            . Log Number: 99-010 Engineering Change Notices (ECNs) 31341 and 31342 install a strobe light in parallel with an audible alarm at the end of panels IH13-P661 and lH13-P662. The installation uses existing                l conduits, and new conduits and supports for the final routing of the conductors to the strobe light.
The strobe light adds an additionalload of 0.03 amps on fire panel power supply, but can still provide four hours of backup power after a loss of the 120-Vac Class IE power source. The strobe lights are visual alarms for the fire detectors. These strobe lights are in compliance with 10CFR50 Appendix R, they are in areas which contain safety related systems or components, and are capable of operating with or without offsite power; The activity does involve a minimal addition of fire hazards by adding wiring for the strobe light, but does not impact the fire load classification of the Main Control Room. The new components meet all applicable design, material and construction standards. Since the circuit is de-energized until a fire is detected, a failure of this new circuit (vould only impact th'e detector circuit. Updated Safety Analysis Report
     . (USAR) Appendix F does not note any requirements for visual alarms when discussing the safe shutdown capability for plant fire areas. The addition of the visual alarms enhances operator
   ,   awareness and reduces the response time for corrective actions which minimizes the fire's impact on safety related systems and components: Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment imponant to safety evaluated previously in the USAR. The postulated failure of the strobe light used to identify a fire could still be identified by the alarm light located at the fire panel and the sounding of the alarm bell. The failure of the strobe light does not impact the operation of the halon suppression which
     - would dump automatically for a floor section fire. Non-safety cables for the strobe lights are routed in conduit and isolated from safety related conductors. Therefore, this change does not create an accident'or malfunction of equipment important to safety of a different type. The
     ' Technical Specifications do not discuss the Fire Protection Program, and therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REPLACEMENT OF THE EXISTING NON-UPS POWER SOURCE FOR TliRBINE BUILDING DIVISION 1 MSL.HI-TEMP MODULES AND ITS ASSOCIATED RELAYS WITH NSPS UPS POWER SOURCE Activity Evaluatedb ECN 29926; USAR Changes 8-259 and 8-298                       Log Number: 99-011 Engineering Change Notice (ECN) 29926 transfers the 120 VAC divisional power of Division 1 Turbine Building Main Steam Line High Temperature Modules and associated relays to the Nuclear System Power System (NSPS) Uninterruptible Power Supplies (UPS) source. The logic
: for Group 1 isolation remains unchanged by this activity. This increases the availability of the -
modules to perform their intended function and reduces the probability of an inadvenent Group 1 isolation occurrence. Calculations indicate that the load increase on the UPS system due to this activity has no adverse effect on the UPS system. The additional load does not have an adverse
      ' affect on circuit breaker requirements, cable sizing, or voltage drop. Changes required to the

y s                                -
4         -
Attachment 2 to U-603276 Page 185 of 252
[                            associated cabling are performed in accordance with the plant design basis. Postulated effects of the failure of the NSPS inverter are high voltage, low voltage, loss of synchronization, or shorts.
LThe resulting effect of these failures is either a transfer to the alternate source or loss of the NSPS
                    .        bus. Effects of the loss of NSPS bus are bounded by analysis performed in the Updated Safety L                            Analysis Report (USAR) Section 15.2.9. Since these modifications are done to the Leak-l1                       l Detection' system circuit loads on the respective NSPS buses only, this activity does not increase      -
the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The only new component added due to this activity is the
                          - cabling and associated relays required to transfer to the existing UPS power source. This is needed to increase availability of the power source. There is no change to the control logic for L the system. Therefore, this change does not create the possibility of an accident or malfunction of H equipment important to safety of a different type than previously evaluated in the USAR. The L                        (direct current sources operating function as described in Technical Specification 3.8.4 is not t                          ' affected. Therefore, this activity does not reduce a margin of safety as defined in the basis for any
                          . Technical Specification;
                           - CLARIFY H202 ANALYZER CHANNEL CALIBRATION AND AUTOMATIC SELF L                          ~ CALIBRATION '
                           . Activity Evaluated: Tech Spec Bases Change BL-99-002                             Log Number: 99-012
[    '
L ' As required by Surveillance Requirement (SR) 3.3.3.1.2, a channel calibration is a complete check
                          . of the instmment loop including the sensor and Main Control Room (MCR) indication and annunciation. Technical Specification (TS) Bases Change BL-99-002 removes the reference to using the automatic drywell .and containment hydrogen and oxygen (H2O2) analyzers' periodic p                          J self-calibration to satisfy channel calibration requirements. The automatic self-calibration does not meet these requirements because it does not include the MCR indication. Clinton Power Station (CPS) Procedure 9437,17," Containment /Drywell Atmosphere H202 Monitoring System
: l.                            Channel Calibration," Revision 36, which previously had been performed every 18 months, has
   .                        l been revised to be performed every 92 days. This activity does not change the TS SR, however,
;                             CPS Procedure 9437.17 must be performed to meet the SR without taking credit for the L                             automatic self-calibration. This more frequent calibration of the entire instrument loop ensures l                             that a malfunction of the hydrogen and oxygen monitoring system is likely to be detected. The H202 monitoring system is an accident mitigation system, it does not act as an initiator for any accident. There is no change to the' system hardware, parameters monitored, or the range that is
: being monitored. In addition, there is no change to the function of the H202 monitors or how
(                             they are used.; Therefore, this activity does not increase the probability or consequences of an l                           ! accident or malfunction of equipment important to safety previously evaluated in the Updated L                             Safety Analysis Report (USAR).' The performance of the channel calibration at the TS required frequency of 92 days provides an opportunity to detect a malfunction of the instrument loop sooner than when the entire loop was checked every 18 months. There are no interconnections
                                                     ~
between the outputs of the monitoring system and inputs of systems important to safety. The system output and use ofthat output is unchanged. Therefore, this activity does not create the
: possibility of an accident or malfunction of equipment important to safety of a difTerent type than l-

W
                                                                                                                            -1
                                                                                                                              ]
Attachment 2              ,     )
to U-603276                     j
, V.                                                                                          Page 186 of 252 previously evaluated, Deleting the automatic calibration from fulfilling TS SR 3.3.3.1.2 means .
that another procedure must be performed. That procedure is.the 18 month calibration which is now being performed every 92 days. Deleting the automatic calibration from the TS Bases results in a full loop. calibration being performed more often than it was previously performed. This
             - activity preserves compliance with TS requirements and the intent of a channel calibration. ;This activity does not reduce a margin of safety as dermed in the basis for any Technical Specification.       -
ADD A BLANK PLATE ON LINES VENTILATION LINES TO ISOLATE RHR ROOM
              - SUMP VENTS -
         ~
L Activity Evaluated:- ECN 31327,31328, and 31329;"                              Log Number: 99-013 USAR Change 8-261:
Engineering Change Notices 31327, 31328, and 31329 make Temporary Modification (TM) 96-093 permanent. This TM installed a carbon steel blank plate between the flanges at the
           ,    interface point between ventilation lines IRF45A-3,1RF62A-3, and 1RF77A-3, and sumps. Each line originates from a flange on top of the sump and connects to the Auxiliary Building Ventilation
             - (VA) system via a mating flange, The blank plates delete the indirect interaction between the VA exhaust system and the Fuel Building Ventilation (VF) system. The blind flanges act as a secondary containment boundary.
             ' Secondary containment, Residual Heat Removal (RHR) equipment, and the Standby Gas Treatment (VG) ' system do not act as initiating events for any evaluated accidents. This change does not impact the operation of the Containment, Auxiliary Building, and Fuel Building floor
             - drain (RF) system pumps or the sump level leak' detection system; Venting of air from the sumps is still possible from the sump cover plate which is not leak-tight. The 3/16 inch thick carbon steel blank plates meet the design, material and constmetion standards for the RF system and VA s        exhaust system. Therefore, this change does not increase the probability of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The small amount of flow lost to the VA system exhaust by implementation of this activity does not adversely affect the VA, VF, and the VG systems operation. This change has no impact on the RF system operation, because the
             ; displaced air from the sumps is released through the cover plate and around the piping penetrating the cover plate. Therefore, this change does not increase the probability of a malfunction of the RHR equipment,''or any other equipment important to safety, previously evaluated in the USAR.
Any gaseous release from the sump pits to the rooms is through the openings around the piping penetrating the sump cover and exhausted by the VF and VG which are the ventilation systems
              ' for the RHR rooms and are designed for this purpose. This change brings the secondary
              - containment into compliance with USAR Section 6.2.3. Therefore, this change does not increase the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The change has no impact on sump operation. The physicalinstallation of the blank plates occurred under TM 96-093. The function of the VF, VA, RHR and RF
: system's has not changed.l Therefore, this change does not create the possibility'of an accident or malfunction of equipment important to safety of a different type. Technical Specification Sections i 3,6.4.1 'and 3.6.4.2 discuss the limits for secondary containment leakage. Its leak tightness is

  ;.                                                                                          Attachmcat 2 to U-603276 Page 187 of 252 required to ensure that the release of radioactive materials from primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis, and that fission products entrapped within the secondary containment structure be treated by the VG J system prior to discharge to the environment. Technical Specification Section 3.6.4.1 arid 3.6.4.3
          . discusses the VG system and states that the system ensures that radioactive materials that leaks from the primary containment into the secondary containment following a Design Basis Accident are filtered and absorbed prior to exhausting to the environment. This change eliminates the vent which was considered to be a secondary containment leakage path, and therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
SWITCHGEAR HEAT REMOVAL CONDENSING UNIT 1VX06CA CONTROL CIRCUIT HOT-SHORT MODIFICATION l Activity Evaluated: ECN 30948                                                    Log Number: 99-014
          ' Engineering Change Notice (ECN) 30948 removes three resistors in the Switchgear Heat Removal Condensing Unit IVX06CA Control Circuit and adds three new resistors in the same electrical circuits but in a different location. Also, this activity adds a one amp fuse to the indication circuits of the condensing unit to protect the control circuit. The new resistors have the   I same rating as the previous resistors, and both the fuse and the new resistors are safety related.
           ' Two resistors are being installed in the 480 VAC Cubicle 1 APl IE, and are environmentally qualified. This modification requires minimal wiring changes. The purpose of this modification is to improve the reliability of the control circuit during a fire in the Main Control Room (MCR).
Following design basis accidents (DBAs), the switchgear heat removal system provides cooling 2 for equipment protection and purging the gas built-up in the battery rooms. Failure of this system is not an event initiator of any DB A defined in Updated Safety Analysis Report (USAR) Chapter           ;
6 or'15. The design modification meets the applicable design, material, and construction standards applicable to this system. The activity does not affect the overall performance of the system. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The activity only makes changes to Division 1 of the switchgear heat removal system._ The modification does not impair the availability of the safety       l l
related systems or' equipment, does not degrade the performance below the design basis, by affecting the environmental, seismic, or separation criteria. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important t'o safety i
            ; previously evaluated in the USAR. ' The design change eliminates the possibility of a short in the safety related switchgear heat removal control circuit. This modification does not alter the function' or operation of the system and does not impact other safety related systems. The components being added meet the same safety and quality design standards as other components installed in this system. ' The fuse is a protective device and protects the circuit in the event of a short during a MCR fire. If the fuse opens prematurely loss cfindication in the MCR occurs.
Loss of control room indication has no affect on the operation of the system. If the fuse fails to open or opens late when needed to perform its protective function, the control circuit would be overloaded by a short. In such a' scenario, the 30 amp control fuse provides the protection of electrical system. Therefore, this change does not create the possibility of an accident or

Attachment 2
* to U-603276 Page 188 of 252..
malfunction of'equipmerit important to safety of a different type. The Technical Specifications do not discuss the switchgear heat removal system. Therefore, this activity does not reduce a margin of safety 'as defined in the basis for any Technical Specification.
1 PROVIDE OFFICE SPACE FOR THE FIN TEAM AT EL. 825' OF CONTROL BUILDING Activity Evaluated: Temporary Modification 99-002 -                            . Log Num'oer:99-015           !
l Temporary Modification 99-002 converts the East side of Control Building Elevation 825' into an office area for the Fix-It-Now (FIN) Team. This temporary modification installs a 480 VAC distribution transformer, desks and computers. There are no design basis accidents discussed in Chapters 6 and 15 of the_ Updated Safety Analysis Report (USAR) affected by this activity. The
_ power feeds are non-safety related and their failure is not identified as an initiator of any accident.
This temporary modification does not impair the availability or degrade the performance below the design basis of any system. This activity does not increase challenges to the safety systems which are assumed to function in the accident analysis be imposing increased or more severe testing requirements, and does not impair system reliability, redundancy, or independence. This
. activity does nct alter the function or operation of any safety systems, and does not degrade or prevent these systems from performing their intended design function. Fire loading in the room has been increased due to addition of combustibles; however, fire loading evaluation concluded 1that fire loading is within the acceptable limits. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety -
previously evaluated in the USAR. ' The modification does not alter the function or operation of arny safety related systems. The components added meet the safety and quality design standards L for their intended service. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The Technical Specifications do not define any requirements potentially affected by this activity.
        ~
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical
 . Specification.
3 DELETE THE HIGH OIL TEMPERATURE TRIP ON THE AUXILIARY POWER TRANSFORMERS Activity Evaluhted: ECN 31330                                                   Log Number: 99-016 Engineering Change Notice (ECN) 31330 deletes the high oil temperature " trip" on the Auxiliary Power Transformers as discussed in the Updated Safety Analysis Report (USAR). No physical work is required as a result of this design change. A review of the electrical schematics concluded that a high oil temperature will provide an alarm function only and does not provide a trip signal.
This change revises the USAR to clarify the potential trips of the Auxiliary Power Transformers
- by deleting the high transformer oil temperature as a potential condition which trips the transformer breakers. This change does not impact the function of the Auxiliary Power Transformers. ? The purpose of the high oil temperature trip is to provide protection of the
; transformers and minimize damage. However, an alarm function provides indication of a potential 4

Attachment 2 to U-603276 Page 189 of 252 malfunction of the transformer without interrupting service. Since this is a document change only
    . to clarify the potential transformer trip signals, this change does not impact any system function'or design basis, and does not impact plant operation. There are no " auto Actions" indicated in the annunciator procedures and the " Operator Actions" include checking the cooling system and shiflirtg'or transferring loads. This documentation change does not impact the alarm function and does not alter any potential failure modes of the alarm function. The transformer high oil
    ' temperature alarm is not involved with any system that mitigates the radiological consequences of an accident. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This change does not perform any hardware change, does not impact any system function or design basir, and does not impact plant operatior). Therefore, this change does not create an accident or malfunction of equipment important to safety of a difTerent type. Technical Specification and Bases 3.8," Electrical Power Systems," describe the offsite power sources required for plant operating and shutdown conditions, but not to the level of detail of the transformer protection or alarm functions. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
MCC BREAKER 1 AP31E. CUBICLE 2D REPLACEMENT Activity Evaluated: ECN 31381; USAR Change 8-254                                 Log Number: 99-017 Engineering Change Notice (ECN) 31381 replaces the 480 VAC Motor Control Center (MCC) breaker I AP31E-2D with a similar breaker that has lower rating. The original breaker is an ITE-Gould Model Number FJ3-B175 rated at 600 Volts AC and 175 amps. This breaker supplies the Shutdown Service Water (SSW) Pump IC. The replacement breaker is an ITE-Gould Model Number FJ3-B150 rated at 600 Volts AC and 150 amps. Failure of this breaker is not an event initiator of the design basis accidents described in Updated Safety Analysis Report (USAR)
Chapters 6 and 15. The new breaker meets the same design, material and construction standards that were applicable to the original breaker. Engineering calculations conclude that the new breaker provides adequate circuit protection and the breaker will not prematurely trip. The modification does not impair the availability of the SSW system'or the electrical system, does not       i degrade the performance below the design basis, by affecting'the environmental, seismic, or              j separation criteria. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The           !
new breaker meets the same safety and quality design standards as other components installed in MCC 1 AP31E. The design change does not alter the function or operation of the electrical and SSW systems and does not impact other safety related systems. The activity does not add any new equipment or components to the systems. Failure modes of the replacement breaker such as inadvertent tripping, premature tripping, failure to open, o'r opening late are the same as the existmg breaker. Therefore, this change does not create the possibility of an accident o'r malfunction of equipment important to safety of a different type. Technical Specification 3.8.9 and 3.8.10 discuss the Distribution Systems during Operating and Shutdown conditions. The margin of safety is defined as " sufficient capacity, capability, redundancy, and reliability."
Electrical calculations indicate that suiricient capacity and capability is provided for this breaker to i

                                                                                                                 )
Attachment 2 -            ,    j to U-603276 Page 190 of 252 perform its safety related function. Redundancy has not been impacted, and reliability remains unchanged since the new breaker is built and installed to the same standards as the original                   .
breaker. Therefore, this activity does not reduce a margin of safety as defined in the basis for any           )
- Technical Specification.-
CHANGE OF MAXIMUM' AVERAGE PLANAR LINEAR HEAT GENERATION RATE                                                -
OPERATING LIMITS FOR GE8B L i
1 Activity Evaluated: Modification NB-033; ECN 31263;                             Log Number: 99-018             l
                         ' USAR Change 8-255
                                                                                                                 ]
                                                                                                                 \
Modification Nb-033 changes the Maximum Average Planar Linear Heat Generation Rage (MAPLHGR) limits in the Core Operating Limits Report (COLR) and 3D Monicore databank for
; reload-6, cycle-7 for some GE8B lattices covering a range of exposure. The purpose of these                    4 changes is to have the peak cladding temperature (PCT) for GE8B type fuel during Design Basis
. Accident Loss of Coolant Accident (DBA LOCA) to be less than 2150 F as opposed to higher                       !
than 2150 FJ For other fuellattices, the DBA LOCA PCT is not higher than 2150 F. The                           j required reduction la the MAPLHGR limit is 2.6% maximum for some specific lattice types                        I covering the exposure from 8.00 to 12.50 GWd/ST.
I The failure modes that act as initiating events for the accidents or transients discussed in Chapters 6 and 15 of the Updated Safety Analysis Report (USAR) are not related to the compliance of
- MAPLHGR limits. Therefore, this change does not increase the probability of an accident previously evaluated in the USAR. The reactor fuel performance is not changed, nor any SSC
 . required to perform safety functions. : Since it requires no change to the application of systems important to safety, this change does not invalidate the initial assumptions for any accident analysis or impact the operation of any equipment assumed to mitigate an accident. Therefore, this change does not increase the probability of a malfunction of equipment important to safety previously evaluated in the USAR. All accidents / transients used as the radiological consequence design bases are evaluated with either a postulated amount of fission product release or conservative coolant activity. This activity lowers the GE8B MAPLHGR operating limit to increase the GE8B margins to the legal limit and thermal-mechanical response limit for LOCA
- PCT and transients, respectively. Hence the fuel clad for GE8B is protected more from the impact of DBA LOCA and transients while the other fuel margin is not affected by this activity.
Since this activity does not affect the original cycle-7 reload analysis or any SSCs, the other fission product barriers are not affected. Therefore, this change does not increase the consequences of an accident or malfunction of equipment important to safety previously evaluated
                ~
in the USAR. Compliance with the new limits during cycle-7 is controlled and achieved using the 7 same plant procedures as before. No component or system is introduced, nor is any component or system changed. Therefore, this change does not create an accident or malfunction of equipment important to safety of a different type. Technical Specifications 3.2.1 and 3.2.3 discuss
: Average Planar Linear Heat Generation Rate (APLHGR) and Linear Heat Generation Rate
 '(LHGR),'respectively. By operating in compliance with the revised MAPLHGR limits, the
' APLHGR for the affected GE8B lattices remain below the new limits. As required by

F      ,
4 Attachment 2 -
!                                                                                          to U-603276
!                                                                                          Page 191 of 252 i.
10CFR50.46, the p~ost-LOCA PCT is limited to 2200 F. By complying with the rs lsed GE8B .
L           MAPLHGR limits, the PCT for GE8B will not be greater than 2150 F during the postulated DBA LOCA. Therefore, this activity does not reduce a margin of safety as defined in the baas for any l
Technical Specification.
L PROVIDETEMPORARY POWER FOR THE STANDBY LIGHTING SYSTEM                                                -
Activity Evaluated: Temporary Modification 99-005                              LogNumber; 99-019 l
Temporary Modification 99-005 provides temporary power to the Standby Lighting System from an existing 100 Amp circuit breaker on the Control Building 480 VAC Riser "C" to provide lighting to Division 1 Diesel Generator Room. The modification requires disconnecting the Standby Lighting cabinet (SLC) 170 from the Class IE power source and connecting it to a non-Class IE source via a temporary cable. The temporary cable is routed on the floor, and is not allowed to be attached to a Class IE division raceway or Class IE component. This temporary
: modification is limited to Mode 4 operations only. Standby Lighting failure is not identified as an event initiator for any design basis accident described in Chapters 6 and 15 of the Updated Safety Analysis Report (USAR). The failure of power feed to the Standby Lighting System has no -                 !
i
         . impact on safety related equipment. This temporary modification does not change, degrade, or
         . prevent actions described or assumed in the accident analysis for mitigating the affect of accidents       ;
         - or transients. This temporary modification provides power to the battery packs during the Bus l Al maintenance outage. Therefore, failure of the SLC would have no impact on safe shutdown of the plant. This temporary modification was installed during Mode 4 and was removed once the
         . Division 1 Bus l Al outage was over. The structural and electrical loadings are within design              ;
limits. The modification does not impair the availability of any system or degrade any system's           j performance below the design basis. Therefore, this activity does not increase the probability or I
consequences of an accident.or malfunction of equipment important to safety previously evaluated in the USAR. The fire loading in the room has been increased due to addition of this cable,              i however, a fire loading evaluation concluded that fire loading is minimally increased and within
                                                                                                                    ,i the acceptable design limits and the fire load rating remains the same. There is no interface
         = between the temporary cable and the safety related system and Class lE power supplies. This temporary modification does not alter the function or operation of the Standby Lighting or the
          . Electrical Distribution System.- Therefore, this activity does not create the possibility of an accident 'or malfunction of equipment important to safety of a different type than previously evaldated.- The Technical Specifications do not define any requirements for the non-safety related power supplies or Standby Lighting System. This design change does not alter Technical Specification requirements for the electrical power systems since there is no impact on the safety-
          . related electrical systems. Therefore, this activity.does not reduce a margin of safety as defined in the basis for any Technical Specification.-
L h

I   s Attachment 2           '
to U-603276 Page 192 of 252 DIVISION III REGULATING TRANSFORMERS J Activity Evaluated: Modification AP-035, Supplement 2;                           Log Number: 99-021 ECN 31059 and USAR Change 8-147
           ;This Supplement' will complete Modification AP-035 by providing a new modular fused
           . distribution panel, e:ectrical schematics and other engineering documents, and by providing appropriate fuse for regulating transformer (l AP78ERT) and for Power Generation Control Complex (PGCC) Main Control Room (MCR) panel fuses. This activity also abandons transformer I AP78E'and the 208/120V AC distribution panel in place.
Partial or complete failure of transformer 1 AP78ERT and loss of some or all of the loads powered from it is not an assumed accident initiator discussed in the Updated Safety Analysis Report (USAR).J Installing appropriately sized fuses for the new modular distribution panel circuits at Motor Control Center ICI and for PGCC MCR panel does not change the functions performed
           - by these circuits : Use of fuses in place of breakers allows proper coordination between connected loads served from the transformer so that a fault on one circuit will not cause the loss of function within other circuits. The design, location, method of operation and potential failures of the new
           - transformer and fuses / modular distribution panel are similar to the original equiprrient. The addition of a regulating transformer adds additional connected load to the Division III Diesel Generator (DG) due to the greater losses in the regulating transformer compared to the noni regulating transformeri Calculations show that the revised load was less than the conservative value listed in USAR Table 8.3-13. Thus, the regulating transformer does not adversely impact t   the continuous rating of the Division III DG. The new equipment is designed to the original plant
           ' de:,ign standards and is capable of supplying 120V power under all plant conditions. None of the changes being made directly affect the availability of the transformer to perform its function.
.          ; Therefore, this activity does not increase the probability or consequences of an accident or
            ; malfunction of equipment important to safety previously evaluated in the USAR. The failure modes of the regulating transformer and fuses are similar to the failure modes of the existing diy
: type transformer and breakerw Thus, there are no new failure modes or malfunctions introduced
           ; by this changec Therefore, this activity does not create the possibility of an accident or.
malfunction of equipment important to safety of a different type than previously evaluated. This activity does not affect any Technical Specification safety limit, limiting safety system setting, or
           . limiting condition for operation. This activity, in conjunction with Modification AP-035, fully restores electrical bus and power distribution systems to an operable condition. Therefore, this
            - activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REVISION OF THE OFFSITE DOSE CALCULATION MANUAL Activity Evaluated: ODCM Revision 18                                         Log Number: 99-022 R/l
            . Changes made to the Offsite Dose Calculation Manual (ODCM) in Revision 18 include following:
change the requirements of the ODCM to reflect implementation of modification PR-04C, include Emodifications arising from an independent review of the ODCM by a consultant, add information

7 I
,, :o               .                                                                       Attachment 2 to U-603276 Page 193 of 252 l         ' governing the Accident Range Monitors, and makes several editorial changes. The editorial changes include clarifying wording, correcting typographical errors, and modifying several unclear formulas; these changes are editorial in nature and do not alter or add any requirements in the "ODCM. The ODCM provides the methodologies and parameters used by Clinton Power Station
: t. assure compliance with the radioactive effluent dose limitations. The applicable accidents            ]
Levaluated in the Updated Safety Analysis Report (USA.R) deal with dose to unrestricted areas           -
from radioactive effluents. The control methodology for the Area Radiation / Process Radiation
          . (AR/PR) system being installed under Modification PR-040 represents an equivalent method of
          . performing the same function as currently specified. In addition, these changes do not alter or
          ' delete any commitment in the ODCM.                                                                       ]
The ODCM methodology is not an assumed initiator of any evaluated accident or malfunction of-equipment important to safety as described in the USAR. 'Nor, does the ODCM provide a mitigatory role to limit radiological releases during any postulated accident. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This activity does not alter any monitoring requirements already in the ODCM or affect operation of any equipment important to safety.1 There are no credible accident or failure modes for equipment important to safety because the ODCM only provides requirements for monitoring effluents and methodology for calculating off site doses. The impact of Modification PR-040 was discussed in Safety Evaluations 98-085 and 98-116 and no concerns were identified.' Therefore, this activity does not create the
           . possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The changes incorporated in this revision of the ODCM are in i          ' full compliance with Technical Specifications for calculating dose to unrestricted areas as a result ofliquid and gaseous effluents. The new AR/PR methodology introduced in Revision 18 is a diffbrent but equivalent way of fulfilling regulatory requirements for control of the AR/PR system.
Therefore, this activity does not reduce a margin of safety as defimed in the basis for any Technical Specification.
            ' IMPROVED TECHNICAL SPECIFICATIONS SR 3.6.3.1.1 BASES CHANGE
             -Activity Evaluated: Techr.ical Specification                                   Log Number: 99-024      i Bases Change BE-99-001 s l             Technical Specification (TS) Bases Change BE-99-001 deletes the requirement to perform a               i "600 F". test in conjunction with a "1200 F" test to determine hydrogea recombiner operability L
from TS Bases for Surveillance Requirement (SR) 3.6.3.1.1.- The "600 F" test is a reduced l
temperature test while the "1200 F" test is the operating temperature test. This activity only deletes the reduced temperature surveillance test that is encompassed by the operating temperature test that is performed at the same time as the test to be deleted. This revises the l*            Technical Specification Bases to be consistent with the previous commitments already reviewed and ~ approved by the Nuclear Regulatory Commission in Safety Evaluation Report (SER) for
            ' Amendment 95aThe hydrogen recombiners do not perform any active function during plant operationi Any failures in the hydrogen recombiner system will not initiate any plant transient or c                             .
l                                        e 3

Attachment 2            ,
to U-603276 Page 194 of 252 accident discussed in the Updated Safety Analysis Repon (USAR). This change does not affect the operability or capability of the hydrogen recombiners to perform their ftmetion to reduce the hydrogen concentration in the' containment following a Loss-of-Coolant Accident. Therefore, this activity does not irisease the probability or consequences of an accident or malfunction of
. equipment imponant to safety previously evaluated in the USAR. The hydrogen recembiners                    -
only function to mitigate the potential consequences of hydrogen generation following a design           -
basis accident. This activity does not introduce any new components or systems into the plant.
Therefore, this activity does not create the possibility of an accident or malfunction of equipment imponant to safety of a different type than previously evaluated. The acceptance criteria for
 . SR 3.6.3.1.1 is that a system functional test will be performed every 18 months. The original Technical Specifications required two functional tests of the hydrogen recombiners: one occurred every eighteen months at the design basis recombiner operating temperatures and a second was conducted at reduced temperatures every six months. This requirement is met by the "1200'F' test, that bounds the "600 F' test. The six month reduced temperature test deletion was approved as documented in'the SER for Amendment 95 which implemented the Improved Technical Specifications. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
. REACTOR RECIRCULATION SINGLE LOOP OPERATION (SLO)
: Activity Evaluated: Modification RRF024; USAR Change 8-107                     Log Number: 99-025 Modification RRF024 incorporates the General Electric (GE) stress analysis resulting from single loop operation (SLO) into the Clinton Power Station (CPS) plant design. CPS submitted a
: license amendment for SLO which was approved by the Nuclear Regulatory Commission, but did not process a plant modification and many documents related to the stress analysis were not updated appropriately. This modification incorporates those documents into the CPS design,'and is a document only change. No hardware change is involvedc .Since this change affect , the
' structural analysis'of the piping, the applicable accident for this change concerns pipe breaks. The GE analyse _s evaluated the changes in thermal movements of piping due to Reactor Recirculation (RR) SLO and piping fatigue analysis. . These analyses were performed in accordance with the same conservative design requirements and limits as the original design. All results were within code allowables. This modification meets the design, material, and construction standards applicable to the systems or equipment being modified.
This activity does not affect overall system performance in a manner which could increase the
: probability or consequences of an accident previously evaluated in the Updated Safety Analysis Repon (USAR). The malfunctions of the RR system evaluated in the USAR are not impacted by the proposed documentation changes.: The activity does not degrade structures, systems or components reliability by deleting or modifying system / equipment protection features, downgrading the suppon system performance necessary for reliable operation of the equipment important to safety, reducing system / equipment redundancy or independence, increasing the.
- frequency of operation of equipment important to. safety, or imposing increased or more severe testing requirements on equipment important to safety. Therefore, this change does not increase

p i
  .                                                                                    Attachment 2 to U-603276 l
Page 195 of 252 the probability of a malfunction of equipment imponant to safety previously evaluated in the USAR. The stress increase due to SLO is analyzed to be within code allowables and does not create the possibility of an accident or malfunction of equipment important to safety of a different type. Technical Specification 3.4.1 discusses Reactor Recirculation and single loop operation.
Since the analyses were performed in accordance with the conservative design limits of Section 111 of the ASME B&PV Code ensuring the integrity of the Reactor Core Pressure Boundary remains unchallenged, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
1 REPLACEMENT OF THE EXISTING NON-UPS POWER SOURCE FOR TURBINE BUILDING DIVISION 2 MSL HI-TEMP MODULES AND ITS ASSOCIATED RELAYS WITH NSPS UPS POWER SOURCE Activity Evaluated: ECN 30032; USAR Change 8-269                                Log Number: 99-026 Engineering Change Notice (ECN) 30032 transfers the 120 VAC divisional power of Division 2 Turbine Building Main Steam Line High Temperature Modules and associated relays to the Nuclear System Power System (NSPS) Uninterruptible Power Supplies (UPS) source. The logic                i for Group 1 isolation remains unchanged by this activity. This increases the availability of the         l modules to perform their intended function and reduces the probability of an inadvertent Group 1 isolation occurrence. Calculations indicate that the load increase on the UPS system due to this activity has no adverse effect on the availability of the UPS system. The additional load does not        i have an adverse affect on circuit breaker requirements, cable sizing, or voltage drop. Changes required to the associated cabling are performed in accordance with the plant design basis.
Postulated effects of the failure of the NSPS inverter are high voltage, low voltage, loss of synchronization, or shorts. The resulting effect of these failures is either a transfer to the alternate source or loss of the NSPS bus. Effects of the loss of NSPS bus are bounded by analysis performed in the Updated Safety Analysis Report (USAR) Section 15.2.9. Since these modifications are done to the Leak Detection system circuit loads on the respective NSPS buses only, this activity does not increase the probability or consequences of an accident or malfunction       ,
of equipment important to safety previously evaluated in the USAR. The only new component                i added due to this activity is the cabling and associated relays required to transfer to the existing UPS power source. This is needed to increase availability of the power source. There is no change to the control logic for the system. Therefore, this change does not create the possibility       ;
of an accident or malfunction of equipment important to safety of a different type than previously      '
evniuated in the USAR. The direct current sources operating function as described in Technical Specification 3.8.4 is not affected. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.

1 Attachment 2               ,
to U-603276 Page 196 of 252 DELETING THE ACTUATOR FOR DAMPER IVOO7Y
  ' Activity Evaluated: ECN 31414, USAR Change 8-262                             Log Number: 99-027 Modification' VO-08 blanked off the air duct in the Off-gas Vault Refrigeration (VO) system that
~ damper IVOO7Y is located in; this renders the ability of the damper to open and close redundant.            -
. This activity deletes actuator, IFZ-VOOO9B, for damper IVOO7Y, The damper and the damper po'sition indication lights are abandoned in place.' Removing the actuator decreases the maintenance work required to maintain the VO system.
The design intent'of the' damper and' associated ductwork was to maintain the pressure in the VO system slightly negative to minimize the potential of radioactive material leakage out of the vault.
Because of the modification, the actuator does not impact the operation of the VO system, and therefore, its removal has no adverse impact upon any transient associated with the operation of the VO system. The equipment directly affected by the VO system is the charcoal absorbers that are a portion of the Off-gas system.' However, the VO system is not required to operate in any -
condition other than normal station operating conditions, has no safety design basis, and is not required for safe shutdown. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment irnportant to safety previously evaluated in the Updated Safety Analysis Report (US AR). This activity removes the actuator for damper IVOO7Yc The failure modes associated with the damper are either failed 'open, closed, or in an intermediate position. The failure modes associated with the duct work are either a duct work breach or duct work blockage. The duct and damper failure modes, have the same affect after the actuator is removed as in the present design. Since the dampers suction is sealed, the operation of
_the damper has no impact upon the operation of the VO system. Actuator removal does not breach the system and does not make the system boundary any more likely to fail. No new equipment is introduced into the plant and does not cause existing equipinent to interact in new waysc Therefore, this change does not create an accident or malfunction of equipment important to safety of a different type than previously evaluated. Technical Specification 3.7.5 discusses the
: release limits associated with.the off-gas system, which to some extent depends upon the
- operation of the vault refrigeration system to meet the release limits. The scope of this activity is -
limited to removing the actuator, this has no adverse impact upon the operation of the VO system or upon the release of radioactive materialsc Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
DEGRADED VOLTAGE CONTROL CIRCUIT UPGRADE FOR CONTROL ROOM HVAC
- SYSTEM -
Activity Evaluated': Modification AP-033 Supp 8; ECN 31369,-                   Log Number: 99-028 USAR Change 8-210 Modification AP-033 Supplement 8 makes the following changes regarding the 120 VAC control circuitry for the Control Room Heating, Ventilation, and Air-Conditioning (HVAC) (VC) Chiller
' Skid Control Panel (0VCl3CA): terminates spare conductors of cables !VC09D and IVC 09B in

9
    .g                                                                                        Attachment 2 i                                                                                             to U-603276 Page 197 of 252 L                       .                    _
l parallel with existing terminated conductors; disconnect and abandon cable 1 VC09J; install new cable;1VC09L to take the place of cable IVC 09J; terminate both circuits of IVC 09L at
;        . OVCl2CA; and terminate the circuit of new cable IVC 09L to the non-regulated distribution
!        : panel.LThis modification does not involve any systems that are design basis accident initiators in 1           Chapter 6 or 15 of the Updated Safety Analysis Report (USAR); it only changes the plant l
         .-support / accident mitigating VC system. This activity does not change the function of the Control     -
Room HVAC Chiller Skid Control Panel control circuit. This modification provides a different power supply in the same division to the control circuit to provide adequate voltage to the components in the circuit. The power source is still a Division 1, Class lE non-regulated distribution panel. Components used in this modification meet the same design, material, testing, separation, and quality' standards as the original components. This modification improves the reliability of the control circuit components to perform their intended design functions. During the installation and testing portions of this modification the plant was in Cold Shutdown ~and
         . appropriate plant procedures, Limiting Conditions for Operation (LCOs), and post-modification testing were adhered to. Modification AP-033 supplement 6 provided the necessary cable work for abandoning cable' IVC 09J and adding cable IVC 09L, so analysis of the cable routing change is addressed in Safety Evaluation 98-160. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously -
evaluated in the USAR. The redesigned control circuit continues to perform its intended design function while adhering to the design standards and practices used in the balance of the circuit.
Thus, this change doer not result in the introduction of an accident of a different type than previously evaluated. Analysis has shown that the non-regulating transformer (0AP54EB) and
: distribution panel are within acceptable loading limits.- Operator control and interface with the circuit remains unaffected. The final configuration associated with this modification does not change the function of the Control Room HVAC Chiller Skid Control Panel circuit or
: components. Therefore, this change does not create the possibility of a malfunction of equipment
                                                              ~
important to safety of a different type than previously evaluated. This change preserves
_ compliance with the Technical Specification requirements and the Bases descriptions. Therefore,
          = this activity does n'ot reduce a margin of safety as defined in the basis for any Technical
          ? Specification.
           . REMOVE CONTROL ROOM VENTILATION SMOKE MODE FROM ORM, ADD TO
           ' USAR SECTION 6.4.5 & APPENDIX E -
Log Number: 99-030
          . Activity Evaluated:1USAR Change 8-270 and -
ORM Change Package 25-5 l
                                                                               ~
This activity makes the following changes: deletes Sections 2.4.3,3.4.3, and 4.4.3," Control Room Ventilation Smoke Mode," from the Operational Requirements Manual (ORM); revises Updated Safety' Analysis Report (USAR) Sections 1.2.2.8.14,6.4, and 7.3.1.L6 to reflect u.at habitability requirements of the Control Room Ventilation system apply only to accident conditions; revises USAR Section 9.4.1 to detail the. testing of the Control Room Ventilation        !
           ' Smoke Mode; removes references to the Control Room Heating Ventilating and Air-Conditioning (HVAC) Smoke Mode, purge mode, and ionization detectors from USAR Section 12.3.3; and L                               ,

Attachment 2           ,
j to U-603276 Page 198 of 252.
: adds a description of the Control Room HVAC Smoke Mode to USAR Appendix E Section
    ,    3.l.2.4. Neither USAR Chapter'6 or 15 take credit for Control Room Ventilation Smoke Mode for prevention or mitigation of any type of accidenti The only event assumed to occur simultaneous to a fire is loss ofoffsite power. . In the event of a loss of offsite power, the Control Room HVAC system' continues to operate. During a failure of the Control Room Ventilation
        . Smoke Mode concurrent to a loss of offsite power and a fire, the safety related functions of              -
temperature control'and protection'of the operators from radiation are maimained. This activity does not affect overall system performance, because it does not physically change the plant or the assumed functional requirements of the system. The Control Room Ventilation Smoke Mode is -
c not required for operability of the control room HVAC system. This change does not result in any change to the way the components of the Control Room Ventilation Smoke Mode are operated; This activity does not impair system reliability, redundancy, or independence.
Therefore, this~ activity does not increase the probability or consequences of an accident or malfunction ofequipment important to' safety previously evaluated in the USAR. No physical changes are made to plant equipment as a result of this activity. This change does not change the
      . way any components of the Control Room Ventilation Smoke Mode are operated. No new components or systems are introduced into ihe plant. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than
      . previously evaluated in the USAR. The Control Room Ventilation smoke detectors and the Control Room Ventilation Smoke Mode are not addressed in the Technical Specifications or Bases. The Control Room Ventilation Smoke Mode is not required for operability of the safety related functions /fechni, cal Specification functions of the Control Room HVAC system.
: Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
RADIATION PROTECTION INSTRUMENT CALIBRATION FACILITY SHIELDING Activity Evaluated: ECN 31466 and USAR Change 8-299                               Log Number: 99-031 The as-built configuration of the Radiation Protection Instrument Calibra, tion Facility (RPICF) is.
       'three 16-inch (nominal) solid concrete. walls, a 20-inch (nominal) thick floor, and 12-inch
      - (nominal) thick ceiling; this differs with the Updated Safety Analysis Report (USAR) description.
USAR Section 12.3.1.10.7 described the RPICF as follows: " Room Shielding consists of three (16-inch (minimum) thick poured concrete walls and 20-inch (nominal) thick floor and ceiling."
This activi+y revises the USAR to match the as-built configuration of the RPICF. The area is not included as a mitigating factor to any accident evaluated in the USAR. The change is not a structural concern, because the as-built configuration of the RPICF is in accordance with the structural drawings and calculations. Radiological Technical Evaluation 99_-003-ED. determines that the shielding changes comply with the RPICF's current Radiation Zone designation. The
: reduction in shielding does not change the dose rates for the areas surrounding the RPICF -
because all surrodnding areas are Radiation Zone A. Changing the density of the walls and the thickness of the ceiling does not affect the fire protection requirements of the room. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This change does not introduce

i-
       .-                                                                                        Attachment 2
     -                                                                                           to U-603276 Page 199 of 252 l'
L any new structure,' system, or component; and therefore, does not introduce any new failure i               ' modes or mechanisms. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the
,                USAR. There are no Technical Specifications or associated Bases that discuss the wall and ceiling characteristicsin the RPICF. Therefore, this activity does not reduce a margin of safety as
              ' defined in the basis for any Technical Specification.                                                     -
DESIGN AND LICENSING DOCUMENTS AFFECTED BY 95 F LAKE WATER (SX)
TEMPERATURE Activity Evaluatedi ECN 27799; USAR Change 8-116                                Log Number: 99-032 This activity is in response to issues identified during the Generic Letter (GL) 89-13, " Service Water System Problems Affecting Safety-Relried Equipment," program review. This activity revises several documents to incorporate a licensing basis 95 F Shutdown Service Water (SX)-
temperature in lieu'of 93 F and editorially corrects various statements. This change affects Updated Safety Analysis Report (USAR) Chapters 1, 2, 3, 5, 6, 7, 8, 9, and 10, various calculations, and the system description for SXc                                                          l Elevated lake temperatures and the associated SX water temperatures are not assumed as indirect -
                ' initiators of any.USAR evaluated accident or design basis event. Numerous supporting calculations' demonstrate the acceptability of the mitigatory function provided by the SX system
               . with water temperatures up to 95 F. A review of the analyses and calculations showed that the
: overall performance of the SX system and station auxiliaries will not cause operations outside of their design limits'nor cause adverse interaction with other systems. This change conforms to the design, material and construction' standards applicable to the SX system. Therefore, this change -
does not increase the probability or consequences of an accident previously evaluated in the USAR. These changes are related to SX flew redistribution and 95 F water temperature which
: affect heat removal from n'umerous coolers. A review of the cooler heat removal performance analyses show that the coolers have adequate' capacity to maintain the temperatures within their maximum operating limit or othetwise provide adequate heat removal. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the U.SAR. The environmental conditions obtained from the analyses which used a design bases lake or SX temperature of 95 F and revised SX flows are bounded by the Equipment Qualification parameters. This activity does not add any new systems,
                                                           ~
                 / structures, or components, nor system interactions. Therefore, this change does not create the possibility'of an accident or malfunction of equipment important to safety of a different type. The
                 ' design basis function of the SX system and its supponed systems have not been altered.
Therefore, this activity does not reduce a margin of safety as defined in th,e basis for any Technical Specification.
l i
          +
l l
1

l l
Attachment 2           ,
to U-603276 Page 200 of 252 CONTAINMENT PRESSURE EPG-RANGE OVERLAP WITH HIGH-RANGE CHANNEL Activity Evaluated: ECNs 31311 and 31312; USAR Change 8-266                    Log Number: 99-033 Containment pressure monitoring utilizes differential pressure cells to measure Containment pressure with respect to atmosphere. Each transmitter senses Containment pressure and transmits this information to an associated recorder located in the Main Control Room. The three recorders associated with each division are required to have overlapping range. Engineering Change Notices (ECNs) 31311 and 31312 implement a new Emergency Procedure Guideline (EPG) range span for containment pressure of 40 to 80 psig in order to provide the required overlap with the high-range channel. This brings the containment pressure monitoring ranges into compliance with Regulatory Guide 1.97 and NUREG-0737 by establishing the low-range at -5 to 10 psig, the high-
. range at 5 to 45 psig, and the EPG-range at 40 to 80 psig.
The Updated Safety Analysis Report (USAR) does not evaluate any accidents affected by the spanning of the EPG-range containment pressure monitoring channels. These channels are for indication only and are for use beyond the postulations of design bases accidents. The-containment pressure monitoring instrumentation performs no active control function, nor does it initiate any accident. This change does not alter the overall system performance, change the system response characteristics, cause the system to operate outside ofits design limits, or cause any adverse system interaction with other systems. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. This change does not change circuit design response and does not present any new challenge to any safety system.
Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This plant change does not change the function or mode of failure of the containment pressure monitoring system. This change does not introduce any new type ofinstrument or equipment or modify existing instrument or equipment that might affect the function or mode of failure of the EPG-range channel. By providing an overlap Operations can make decisions or take actions without interruption of the high limit of the high-range channel or the low limit of the EPG-range channel. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to         ,
safety of a different type. Technical Specification 3.3.3.1 discusses the low-range channel and the high-range channel, however, the EPG-range channel was not a part of the original licensing basis.
The EPG-range channel was a commitment made later as a result of the Three Mile Island-2 accident. This change does not impact the requirements in the Technical Specifications, and              i therefore, does not reduce a margin of safety as dermed in the basis for any Technical Specification.
1 I
U

       ~
.i                                                                                             Attachment 2 to U-603276 -
Page 20I of 252
          -ADDITION OF APPENDIX R MANUAL ~ ACTIONS FOR FIRES OUTSIDE THE MAIN CONTROL ROOM '
Activity Evaluated:- CPS Procedure 1893.04 R/8 and                                  Log Number: 99-034 USAR Change 8-267 This change revises Clinton Power Station (CPS) Procedure 1893.04, " Fire Fighting," to provide
         ; guidance to Main Control Room (MCR) operators relative to achieving Safe Shutdown of the plant in the event of a fire. Updated Safety Analysis Report (USAR) Appendix E," Fire Hazards
         . Analysis," states'that for a fire anywhere in the plant, Safe Shutdown can be accomplished.
USAR Appendix F," Safe Shutdown Analysis (SSA)," demonstrates that for a fire in'any single
         - plant area, at least 'one method exists that is free of fire damage to achieve and maintain a safe shutdown condition. This procedure revision does not introduce any new fire hazards that would
         ; affect the Fire Hazards Analysis or SSA. The procedure revision does not change the design, function, operation, or test of any equipment or systects affected. This activity clarifies and lists available manual actions to assure the timely operation, or preclude or mitigate the inadvertent operation of the equipment or systems involved due to a fire. This activity does not alter the way equipment is operated, it provides guidance for available manual actions that may be taken for a fire in a specific fire area / fire zone. Therefore, this change does not increase or create the -
probability or consequences of an accident or malfunction of:quipment important to safety. The Technical Specifications do not discuss the CPS Fire Protection Program. Therefore, this procedure revisbn does not reduce a margin of safety as defined in the basis for any Technical -
          - Specification.
REPLACEMENT OF OBSOLETE BUSSMAN FUSES IN ELECTRICAL PENETRATION CIRCUITS Activity Evaluated: ECN 31432; USAR Change 8-268                                    Log Number: 99-035 Engineering Change Notice (ECN) 31432 updates the design drawings for electrical penetration
: circuits to allow the use of a Gould A2KR fuse in place of the Bussman KWN fuses. Bussman no longer manufactures KWN fuses and the replacement fuses are very similar to the original fuses, except they are provided by a different vendor and have a different fuse catalog number. Theu fuses provide protection for the electrical penetration and a failure of these fuses does not initiate any accidents described in the Updated Safety Analysis Report (USAR). The modification meets the applicable design, material, and construction standards applicable to the system and equipment being modified.lThe replacement fuses have the same rating, phy+nt dimension, interrupting rating, and current limits as the original fuses. The activity does v      iffect the overall system pGrmance or protective function. This modification does not auer the function or operation of l any safety system, and does not degrade or prevent these systems from performing their intended design function. Therefore, this change does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the USAR. No new components are added.- The replacement fuses are similar to the original fuses and meet the same safety arid quality design standards as other components installed in the applicable circuits.

Attachment 2
* to U-603276 N                                                                                           Page 202 of 252 Technical Specification 3.3 discusses the electrical power system, but does not discuss electrical
_     penetrations nor does it discsss fuses to this level of detail. This design change does not alter the Technical Specification requirements for the electrical systems. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
                                                                                                                        )
          - HYDROGEN IGNITION SYSTEM USAR CHANGE                                                                      -
Q Activity Evaluated: USAR Change 8-256;                                          Log Number: 99-036 Updated Safety Analysis Report (USAR) Change 8-256 removes the USAR requirement to compare surveillance test data to pre-operational test data in favor of using acceptance criteria,
: stated within Clinton Power Station Procedure 9367.04, " Hydrogen Igniter Current Test," for voltage and current levels in order to accurately determine Hydrogen Igniter operability. Pre-              ,
operational test data was only taken for the situation where the igniter supply voltage was 120 LVac. This would be acceptable if all of the igniters could be maintained at 120 Vac all of the time.
However, normal variances in voltage are both expected and acceptable. The Hydrogen Ignition (HI) System is an accident mitigation system, and therefore, the hydrogen igniters are not accident initiators. This change does not adversely impact the system's response to an accident. The HI system continues to meet its design function. The change continues to conserutively assure that the individual igniters are capable of attaining a minimum of 1700 F. By specifically stating the criteria in a procedure that supports the USAR requirements, the potential for false calls due to -
interpretation errors are eliminated.- Clarifying this criteria does not adversely impact system operability, nor dose it pose any degradation to the HI system. 'Also, this USAR change does not -
reduce system redundancy or independence. Therefore, this activity does not increase the probability or consequences ofan accident or malfunction of eguipment important to safety previously evaluated in the USAR. No new equipment, systems, or methods of operation are being added. The actual test methodology remains unchanged. The igniters continue to operate
        - as before Therefore, this activity'does not create the possibility of an accident or malfunction of
           - equipment important to safety of a different type than previously evaluated. Technical
          ' Specification 3.6.3.2 and its associated Bases discuss the HI system. Both the Technical Specification and the associated Surveillance Requirements are based upon the fact that the igniters are capable of achieving at least 1700 F. When determining the setpoints for the
: procedure, all assumptions were. applied conservatively in order to assure that the 1700 F is achievable in all conditions. Therefore, this activity does not reduce a margin of safety as defined
: in the basis for any Technical Specification.
           . SHUTDOWN SERVICE WATER SYSTEM VALVE CONTROL CIRCUIT CHANGES 4 Activity Evaluated:L ECN 31405; ORM Change 25-6;                            Log Number: 99-037 R/l USAR Change 8-271 md 8-337
                                                                                                                        )
Engineering Change Notice (ECN) 31405 reconfigures Shutdown Service Water (SX) system controllogic and valve positions in order to establish an alternate 1SX0lPA pump minimum flow path while maintaining the required flow. This is accomplished by defeating the auto-open logic n   -
                                                                                                                        )

  >4                                                                                      Attachment 2 to U403276 Page 203 of 252 to the existing minimum flow valve and providing auto-open to two new' minimum flow bypass valves.: The SX system is an accident mitigating system. Most accidents require the operation of the SX system for successful accident mitigation; however, the SX system does not initiate any
       - accident described in the Updated Safety Analysis Report (USAR). This activity does not; degride the_ Residual Heat Removal or SX system because a suitable path for the ISX01PA pump minimum flow is established and flow rates are maintained within design parameters. This activity       -
       ' does not disable any automatic SX system initiations in response to an accident or Service Water
        ' low pressure signal. The new minimum flow path is established with components meeting the same design, material and construction standards as the original flow path. The proposed flow
       , rate is within the manufacturer's allowables and results in no degradation to the pump. The plant -
design allows the complete loss of one SX division while still safety shutting down the plant.
Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. ' Failure modes which could be introduced by these changes are failure of one of these new valves to open; degradation of the basket strainer; reduced efficiency to the heat exchanger form erosion, corrosion, or silting; degrading capability of the diesel generator cooling water system to maintain the block and lube oil heating to facilitate fast engine stans. None of these failures would result in the initiation of an accident. No new equipment, functions, or flow paths are being added, and the systems still operate within their design limits. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluatsd. The Technical Specifications discuss Shutdown Service Water and auxiliary loads provided by SX.' A reduction in the margin of safety would be tied to accident conditions or system degradation resulting from operation to support testing or on the ability of the diesel '
generator to load the bus within the specified time frame. Failure to establish the minimum flow path does not directly affect the ability of the SX system to provide design basis flow. Therefore,-
this activity does not redu'ce a margin of safety as defined in the basis for any Technical Specification.
H202 MONITOR INSTRUMENT ACCURACY Activity Evaluated: USAR Change 8-273 '                                        ; Log Number: 99-038 Updathd Safety Analysis Report (USAR) Change 8-273 revises USAR Section 7.6.1.10.11.2.b t and USAR Appendix D Section II.F.1.f stated instrument accuracy for the Containment and Drywell Hydrogen / Oxygen Concentration Monitor Indicators from 1.0% of full scale to 1.0% by volume.; In addition, this USAR change deletes the reference to " indicated" when stating the hydrogen concentration. to activate Combustible Gas Control System equipment from USAR Sections 6.2.5.1.1.b and 6.2.5.1.3.3. The operation or failure of the Containment and Drywell Hydrogen / Oxygen Concentration Monitoring System is not an initiating event for any transient or accident. The Containment and Drywell Hydrogen / Oxygen Concentration Monitoring System functions to provide the capability of monitoring and indicating normal operating and                   ,
post-accident hydrogen / oxygen concentrations in the drywell and containment. This activity does not change, degrade, or prevent actions described for hydrogen mitigation following a
         - Loss-of-Coolant 'Acc _ uent. This activity does not affect the operability, capability, or introduce l
l I
1

Attachment 2
* to U-603276.
Page 204 of 252 any new failure modes of the system. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated
    ;in the USARL This activity'does not change the operation, function, or reliability of this system.
This activity does not introduce any new components or systems into the plant. Therefore, this activity does not create the possibility of an accident or malfunction of equipment imponant to            J safety of a different type than previously evaluated. ' Technical Specification Basis 3.6.3.1, 3.6.3.2     l and 3.6.3.3 state that calculations confirm that when the hydrogen mitigating systems are actuated         l Lin accordance with plant procedures, the peak hydrogen concentration in the primary containment             {
remains less than 4% by volume. This change in instrument accuracy from 1.0% of full scale to
{'
l.0% by volume does not affect the operability or capability of the Containment and Drywell Hydrogen / Oxygen Concentration Monitoring System to provide the hydrogen concentration indication to provide the hydrogen concentration indication to initiate ' activation of the l
Combustible Gas Control System to contral hydrogen concentration below 4%. Therefore,'this                 i activity.does not reduce a margin of safety as defined in the basis for any Technical Specification.
DE-ENERGIZING AND LOCKING IN THE CLOSED POSITION CONTAINMENT ISOLATION VALVES 1FP051, IFP054,' IFP078, AND IFP079 Activity Evaluated: ECNs 31386,31480, and 31481;                                . Log Number: 99-039 USAR Change 8-275 Engineering Change Notices (ECNs) 31386,31480, and 31481 de-energize and lock closed containment isolation valves IFP051 and IFP054, and drywell isolation valves IFP078 and IFP079. The design change also adds a locking device on the actuator and valve handwheel to prevent manual opening of the valves. This activity changes these valves to manually-operated
    . valves and the system valve lineup has been changed. The Updated Safety Analysis Repon (USAR) does not consider the operation or thilure of the Fire Protection (FP) System as an initiating event for an accident scenario. Tim safety function of the FP Containment /Drywell Isolation Valves is to provide a Containment /Drywell Isolation Boundary by being closed. By de-energizing and locking the valves, the modification prevents insdvertent opening and ensures that the valves are normally closed during all modes of operation, thereby performing their safety function. The valves presently require remote operation by an Operator in order to be opened.
This modification will require an Operator to either close the breaker at the motor control center         !
or to manually open the locked closed valve. This results in an additional s;ep to open the valves.
Ilowever, the time required to.open the valves does not adversely impact personnel's ability to combat a fire in the affected areas. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The logic and circuitry of these valves do not interlock with other equipment other than the interlock to close these valves during the initiation of the containment isolation signal.
This activity; simply converts the valve's safety position to coincide with its normal position and does not add any new structure, system, or component.1 Any concern that the valves may fail to close is no longer a concern since they become manually-operated locked-closed valves and are              ;
administratively controlled. Since the valves are maintained closed and operator action is already
    . a requirement for operation, no new or more probable failure modes have been introduced.
i l
i 1

p l    s
  .'                                                                                         Attachment 2 to U-603276 Page 205 of 252 Therefore, this change does not create the possibility of an accident or malfunction of equipment l           important to safety of a different type than previously evaluated. This change does not impact the
           . ability to maintain containment integrity, therefore, this activity does not reduce a margin of safety
         - as defmed in the basis for any Technical Specification.
UPDATE OF FIRE PROTECTION SAFE SHUTDOWN EQUIPMENT LIST                                                  -
: Activity EvaluatediPSAR Change 8-274-                                             Log Number: 99-040 Updated Safety' Analysis Report (USAR) 8-274 addresses the completion of the updated Safe Shutdown Equipment List (SSEL) for the Safe Shutdown Analysis (SSA), USAR Appendix'F.
Based upon existing licensing documents, design basis documents, analyses, and calculations, plant systems were selected which fulfilled the safe shutdown performance goals outlined in 10 CFR 50 Appendix R.- USAR Appendix F demonstrates that for a fire in any single plant fire area, at least one method exists that is free of fire damage to achieve and maintain a safe shutdown condition.:. In response to Condition Report 3-97-06-310-0,10CFR50 Appendix R Safe
!            Shutdown Validation Project identified the Safe Shutdown conclusions of USAR Appendix F as being in compliance with 10 CFR 50 Appendix R, with identified exceptions. This activity
         . updates the USAR Appendix F SSEL based on the activities accomplished under CR 3-97              310-0, and assures that there is no compromise in the capability to perform a Safe Shutdown.
This activity does not involve a change to the design, function, operation, or test of any .             4 equipment or systems. This activity involves the documentation of existing safety systems and
           . components required for safe shutdown and has no affect on the environmental, seismic, or separation criteria of the affected systems. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment previously evaluated in the USAR.HThis activity assures that safe shutdown systems and components are identified and are bounded by the existing SSA. This activity does not involve any new equipment or a change
: to the design, function, operation, or test'of any existing equipment or systems. Therefore, this
           . activity does ut increase the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The Technical Specifications do not discuss the .Clinton Power Station Fire Protection Program. 10 CFR 50 Appendix R establishes design
            . limitations for the Fire Protection System. This change assures that safe shutdown systems and L             components required to mitigate the consequences of a design basis fire are identified and comply
             'with the requirements of 10 CFR 50 Appendix R. Therefore, this activity does not reduce a b          xargin of safety as' defined in the basis for any Technical Specification.
L             RECONFIGURATION OF 120VAC CIRCUITS - CONTAINMENT MONITORING (H202)
            . SYSTEM
:             Activity Evaluated: Modification AP-047 Supplement 4 Log Number:- 99-041 Modification AP-047, Supplement 4 reconfigures the Division 1 and Division 2 H202 Containment danitoring Systems to ensure voltages at the system's electrical devices are within the allowable operating rangeL The revision to the circuit configuration places spare conductors 9

Attachment 2             ,
to U-603276
                                    .                                              Page 206 of 252 in parallel with existing conductors. Paralleling of existing wiring enhances the component reliability by lowering the effective impedance of the feeder cable, thus reducing the voltage drop and increasing voltage available at the components. In addition, a 0.1 Amp fuse is added in each
  ' Division of a separate circuit between the Silicon Controlled Rectifier and a device located in the Containment Building. Adding a 0.1 Amp fuse is in accordance with the Updated Safety Analysis Report (U_S AR) description. The fuse is sized to protect the penetration and will now contain the required redundant coritainment penetration fuse protection. This activity does not change the function or method of operating any of the circuits. This modification does not change any
   . systems that are design basis accident initiators discussed in Chapters 6 and 15 of the USARc Wiring changes are designed to the same standards of separation and are bounded by existing :
seismic design and ' safe shutdown requirements. Radiological effects or radiological release mitigation paths are not affected. Therefore, this' activity does not increase the probability.or .
consequences of an accident or malfunction of equipment important to safety previously evaluated '
  - in the USAR. These circuits are part of post accident monitoring systems, and these changes improve the available voltage at the devices, thereby improving their function to act as monitors after a ptential accident., The final configuration associated with this modification does not -
change the rision or method of operating the affected circuits or components. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. Applicable Limiting Conditions for Operations, procedures, installation specifications and post-modification testing were adhered to, ensuring the modification was installed in accordance with the design and within plant operational -
constraints. No setpoint or design margins are adversely affected by this design change.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical
  . Specification.-
SHUTDOWN SERVICE WATER SYSTEM VALVE CONTROL CIRCUIT CHANGES Activity Evaluated: ECN 31406; ORM Change 25-9;                                Log Number: 99-042 and USAR Changes 8-276 and 8-337-Engineering Change Notice (ECN) 314% reconfigures Shutdown Service Water (SX) system controllogic and valve positions in order to establish an alternate ISX0lPB pump minimum flow path while maintaining the required flow. This is accomplished by defeating the auto-open logic to the existing minimum flow valve and providing auto-open to two new minimum flow bypass valves. The SX system is an accident mitigating system. Most accidents require the operation of the SX system for successful accident mitigation; however, the SX system does not initiate any accident described in the Updated Safety Analysis Report (USAR). This activity does not degrade the Residual Heat Removal or SX system because a suitable path for the 1SX0lPA pump minimum flow is established and flow rates are maintained within design parameters. This activity does not disable any automatic SX system initiations in response to an accident or Service Water low pressure signal. The new minimum flow path is established with components meeting the same design, material and construction standards as the original flow path. The proposed flow rate is within the manufacturer's allowables and results in no degradation to the pump. The plant design allows the complete loss of one SX division while still safety shutting down the plant.

4 j                                                                                         Attachment 2 to U-603276 Page 207 of 252 l Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. Failure modes 1which .could be introduced by these changes are failure of one of these new' valves to open; degradation'of the basket strainer; reduced ' efficiency to the heat exchanger form erosion,
: corrosion,.or silting; degrading capability of the diesel generator cooling water system to maintain the block and lube oil heating to facilitate fast engine starts. None of these failures would result in .
        - the initiation of an accident.1No new equipment, functions,' or flow paths are being added, and the systems still operate within their design limits. Therefore, this change does not create the -
possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The Technical Specifications discuss Shutdown Service Water and auxiliary loads provided by SX.1 A reduction in the margin of safety would be tied to accident conditions or
: system degradation resulting from operation to support testing or on the ability of the diesel
         > generator to load the bus within the specified time frame. Failure to establish the minimum flow path does not directly affect the ability of the SX syst'em to provide design basis flow. .Therefore,
         ~t his activity does not reduce a margin of safety as defined in the basis for any Technical
          ' Specification. -
         ._ REVISION TO RCIC PRESSURE-TEMPERATURE TIME HISTORY FOR CLASS 1 PIPING
           - AND COMPONENTS Activity Evaluated; Calculations (See SE for Listing);                           Log Number: 99-043
                                      . USAR Change 8-300 This change updates the Updated Safety Analysis Report (USAR) to reflect the results of analysis performed for the Reactor Core Isolation Cooling (RCIC) system. This change corrects a discrepancy found between the number of RCIC injections into the vessel during composite loss of feedwater events versus the postulated number of events analyzed in the piping and component fatigue analyses.- It also corrects a discrepancy found where the flow rate through the RCIC injection line is greater than what was evaluated in the fatigue evaluations. This activity also revises the pressure-temperature time history for the injection portion of the RCIC system. Also,
            ''this activity revises calculations associated with stress reports for the .RCIC class 1 piping, primary containment penetration IMC-42, Reactor Pressure Vessel Vent and Head Spray assembly,' and bulkhead penetration 1B.H-10. This change only affects the integrity of the RCIC l system piping and penetrations. A RCIC line break is the only accident previously evaluated in -
L the USAR that' pertains to this change. For the RCIC spray piping and penetrations, revising the transient flow rates and number of cycles causes an increase in secondary stresses as well as cumulative fatigue usage. However, the stress report results reveal that all stresses and cumulative usage are below ASME Code allowables and USAR limits. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR; The only impact this change can have is
           - the increased potential for crack initiation or crack propagation on the RCIC piping or                 ,
I penetrationss However, this potential increase is offset by updated transient analysis ano an
            ! updated component lifetime which would dictate the need for component replacement. Since, this change does not add any new structures, systems, or components and does not change the

Attachment 3             ,
to U-603276 '
L.                                                                                   Page 208 of 252 function of the RCIC system, this change does not create the possibility of an accident or           '
malfunction of equipment of a different type than previously evaluated. Technical Specifications 3.4.11,3.5.3.3, and 3.5.3.4 discusses Reactor Coolant System (RCS) and RCIC flow rates.
Portions of the RCIC head spray system are included in'the RCS pressure boundary.- Heatup and I
   ~ cooldown rates specified in the Technical Specifications' correlate to those used in fatigue -
evaluations.' However, heatup'.and cooldown rates are not revised as a result of the change. Nor are any temperature or pressure' limits listed in the Technical Specifications being affected by the change; only the number ofinjections. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.-
DE-ENERGIZING AND LOCKING DRYWELL ISOLATION VALVES ICY 020 AND 1CYO21.IN THE CLOSED POSITION -
Activity-Evaluated: ECNs 31400,31460, and 31461;                 _
Log Number 99-044 USAR Changes 8-275 and 8-278; and
                            - ORM Change 25-8 Engineering Change Notices (ECNs) 31400. 31460, and 31461 change the normal position of
   - Drywell Isolation Valves (DIVs) ICY 020 aad ICYO21 in response to Generic Letter 89-10.. This change de-energizes the valves and locks the power switch in the closed position. A locking device has been added to the actuator and valve handwheel to prevent manual opening of the .
valve.1 Bypass switches have been added at the motor control center (MCC) to bypass ' alarms and indications for the associated valves to prevent false alarms and indications while the valves are de-energizedc The Updated Safety Analysis Report (USAR) does not discuss any accident
: initiated by the Cycled Condensate (CY) system. The safety function of the CY DIVs is to provide a Drywell Isolation Boundary by being closed in the required time subsequent to the
   - initiation of the Drywell isolation signal.-. The safety function changes from active to passive. By de-energizing and locking the valves, the m'odification prevents inadvertent opening and ensures that the valves are normally closed during all modes of operation, thereby performing their safety function.' This modification requires an Operator to either close the breaker at the motor control
    . center or to manually open the locked closed valve. . This results in an additional step to open the valves. However, there is adequate time to open the valves at the MCC switch or manually manipulate the valve. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
The logic and circuitry of these valves do not interlock with other equipment other than the interlock to close these valves during the initiation of the containment isolation signal. This
     ' activity simply converts the valve's safety position to coincide with its normal position aiH does
   = not add any new stmeture, system, or component. Any concern that the valves may fail to close              j
   - is no longer a concern since they become manually-operated locked-closed valves and are administratively controlled. Therefore, this change does not create the possibility of an accident
: or malfunction of cquipment impartant to safety of a different type than previously evaluated.
This change does not affect the performance or the operability of either the Cycled Condensate System or Containment isolation. Technical Specification 3.6.5.3 discusses Drywell isolation.
   - Sealing the valves closed changes the surveillance requirements of the valves, but, the proposed J
m

lL Attachment 2 to U-603276 Page 209 of 252 changes are within the Improved Technical Specification bases. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
DIVISION 2 CONTAINMENT MONITORING TEMPERATURE CONTROLLER RACK o        POWER-SUPPLY CHANGE Activity Evaluatedi ECN 31426                                                  Log Number: 99-045 l
Engineering Change Notice (ECN) 31426 changes the power supply for the Division 2 ~
       . Containment Monitoring Temperature Controller Rack from the 120 VAC power supply for panel
                         ^
lH13-P868 to the 120 VAC power supply for the Air Compressor controls. This involves rewiring the Air Compressor Control Power and removing a fuse from the panel. Adequate isolation, separation, and independence is provided in accordance with the Updated Safety.
        ' Analysis Report (US AR), industry codes and standards, and regulatory guides to preclude any l         adverse affect on the Class IE AC Power Distribution System. Adequate breaker-fuse coordination is provided to inhibit an electrical fault from propagating through the power distribution system. The Containment Monitoring System is used in normal and accident conditions to monitor the hydrogen concentration in containment and the drywell, and provide operators with alarms and indications. The new components are designed, manufactured, tested, and installed in accordance with USAR requirements, industry codes and standards, and approved plant procedures to ensure system reliability, and preclude any adverse affect on the failure modes of the Containment Monitoring system. - There are no new system interactions introduced by the changes and the Class IE AC Power System continues to perform as required. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.- The changes do not affect the
         . design intent 'of the Containment Monitoring System or the Class 1E AC Power System. This
        - change does not introduce any new system interactions. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The Technical Specifications do not discuss any design limitations for
: the Containment Monitoriug System or Class IE AC Power System. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
DIVISION 1 CONTAINMENT MONITORING TEMPERATURE CONTROLLER RACK
         ' POWER SUPPLY CHANGE -
Activity Evaluated: ECN 31425                                                 Log Number: 99-046 Engineering Change Notice (ECN) 31425 changes the power supply for the Division 1 Containment Monitoring Temperature Controller Rack from the 120 VAC power supply for panel 1H13-P867 to the 120 VAC power supply for the Air Compressor controls. This involves L           rewiring the Air Compressor Control Power and removing a fuse from the panel. Adequate L           isolation, separation, and independence is provided in accordance with the Updated Safety
[          ~ Analysis Report (USAR), industry codes and standards, and regulatory guides to preclude any adverse affect on the Class IE AC Power Distribution System. Adequate breaker-fuse M

Attachment 2         ,
to U-603276 Page 210 of 252 coordination is provided to inhibit an electrical fault from propagating through the power distribution system. The Containment Monitoring System is used in normal and accident conditions to monitor the hydrogen concentration in containment and the drywell, and provide operators with alarms and' indications. The new components are designed, manufactured, tested,
  , and installed in accordance with USAR requirements, industry codes and standards, and approved plant procedures to ensure system reliability, and preclude any adverse affect on the failure modes of the Containment Monitoring system. There are no new system interactions introduced by the changes and the Class IE AC Power System continues to perform as required, Therefore, this
  . change does not increase the probability or consequences of an accident or malfunction _of .               l equipment important to safety previously evaluated in the USAR. The changes do not affect the design intent of the Containment Monitoring System or the Class IE AC Power System; This change does not introduce any new system interactions. Therefore, this change does not create
                    ~
the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The Technical Specifications do not discuss any design limitations for the Containment Monitoring System or Class IE AC Power System. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
l INSTALL RESTRAINING MECHANISM ON CB AND DG BUILDING ROLLING STEEL DOORS SDl-55,61,63,64,66, AND 67 Activity Evaluated: ECN 31452 and USAR Change 8-280                               Log Number: 99-047
(
  ' Engineering Change Notice (ECN) 31452 adds restraining mechanisms to the Control Building (CB) and Diesel Generator (DG) Building rolling steel doors to prevent inadvenent opening of the rolling steel doors due to fire induced electrical circuit failure. This restraining mechanism results in the doors being treated similar to other locked fire doors. The three hour fire rating of     )
the barriers is maintained by keeping the doors closed and restrained. This change does not affect        )
   . any design basis accidents discussed in the Updated Safety Analysis Report (USAR). Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR.. Restraining the doors and the administrative controls along with posted signs on -
the doors ensures that the probability and consequences of a malfunction of the DG Division 1,2, and 3 do not increase beyond those already evaluated in the fire hazards analysis. There are no new failures introduced, because the doors are in the fail safe position. These doors are maintained closed, except when required to open for maintenance purposes. When open for this purpose, the compensatory measures established in the fire protection program and implemented
  . in station procedures would be in effect No structures, systems, or components are introduced by this activity. This change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The Technical
: Specifications do not discuss the rolling steel doors or the requirements for fire protection.
: Therefore, this activity does not reduce a margin of safety as defined in the Technical
  ~ Specifications.

  .                                                                                   Attachment 2 to U-603276 Page _.. of 252 LIFTING RCIC CABLE TERMINATIONS TO MINIMIZE POTENTIAL FIRE-INDUCED FAILURES Activity Evaluated: ECN 31486; USAR Change 8-284                                Log Number: 99-048 Engineering Change Notice (ECN) 31486 lifts four cable terminatims inside the local Reactor Core Isolation Cooling (RCIC) Turbine Control Panel to eliminate a potential fire induced fault of the RCIC system governor. Since the conductors lifled provide signals to the Transient Test panel and the Transient Test points are no longer used, the points were spared. RCIC is designed as an accident mitigation system, but can create an increase in reactor coolant inventory transient ifit were to spuriously start. This accident identifies two potential initiators for inadvertent RCIC pump start; operator error and false start signa!. This modification does not increase the probability of operator error, because it does not introduce any change to the operator interface with the RCIC system. The Transient Test points are passive and only accept signals for the RCIC governor. No other changes are made that could result in a false start signal for the system.
Thus, eliminating the conductors has no affect upon the operation of the governor. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the Updated Safety Analysis Report (USAR). This change does not impact the design, function, or method of performing the function of the RCIC system, and the moi W                 'no adverse impact upon the Fire Protection Program. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the US AR. Since the Transient Test points being spared were not previously used, sparing these points does not result in any procedures changes. This activity does not introduce any new credible failure modes, nor add any new structures, systems or components. Therefore, this change does not create an accident or malfunction of equipment of a different type than previously evaluated. The changes do not alter the control logic, alarms, or operation discussed in Technical Specifications 3.3.5.2 and 3.5.3. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REMOVAL OF CALGON SCALE INHIBITING EQUIPMENT AND ASSOCIATED PIPING Activity Evaluated: ECN 30153; USAR Change 8-289                                Log Number: 99-049 Modification CL-007 abandoned the Circulating Water (CW) scale inhibitor injection equipment in place. The equipment was not removed when Modification CL-007 implemented as a contingent for possible failure of the new equipment. Temporary Modification 97-036 removed some of the tubing and installed piping associated with the CALGON injection scale inhibiting equipment. Engineering change Notice (ECN) 30153 removes the remaining obsolete equipment and revises drawings and other documentation to reflect the absence of this equipment.
Since the equipment has been disconnected both mechanically and electrically from the CW system, it does not perform any function that could affect the probability of an accident. The removal of the non-functioning CW scale inhibitor injection equipment does not change, degrade, or prevent actions described or assumed in the accident analysis. Therefore, this change does not f
l I
L

h                s Attachment 2             ,
to U-603276 Page 212 of 252 increase the probability or consequences of an acciden_t previously evaluated in the USARc The CW system is not required to perform or to assist in performing any nuclear safety functions.
     . Also, the removal of the obsolete equipment does not affect any other systems or equipment that are considered importan(to safety. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR.L This activity does not introduce any new credible failure modes or any new structures, systems, or components. Therefore, this change does not create the possibility of an accident or
     . malfunction of equipment important to safety of a different type. The Technical Specifications do not discuss the CW scale inhibitor injection system, CW corrosion monitoring, or CW chemistry.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specifications.
   - ' RHR HEAT EXCHANGERS SHELL SIDE MINIMUM DESIGN TEMPERATURE CHANGE FROM 40*F TO 32 F ~
m Activity Evaluated: ECN 31255, Part 1; USAR Change 8-286                       Log Number: 99-050 Engineering Change Notice (ECN) 31255, Part 1,~ revises the minimum design temperature of the Residual Heat Removal (RHR) heat exchangers shell side from 40 F to 32 F. The original General Electric (GE) American Society of Mechanical Engineers (ASME) code calculations had been certified for 32 F on both the shell and tube side of the heat exchanger. The tube side is rated at 32 F. ' When the GE specification was provided to Illinois Power, the minimum -
temperature for the shell side was changed to 40 F. GE review / analysis shows that lowering the
     . shell side of the RHR heat exchanger to match the tube side does not impact the function of the heat exchangers, pressure integrity, structural adequacy, and continues to satisfy the ASME code requirements.'
Chapter 15'of the Updated Safety Analysis Report (USAR) identifies Failure of RHR Shutdown Cooling as a design basis accident. Since the' original design documentation establishes that the shell side is qualified to 32 F, this change does not increase the likelihood of a Loss of Coolant Accident. (LOCA) or loss of shutdown cooling. Reducing the minimum temperature of the shell side of the RHR heat exchangers does not adversely impact any other system, because the only
     ; systeminterface already has a minimum design temperature of 32 F. This change does not reduce the reliability of the RHR heat exchangers to supply the required cooling capability to limit Containment pressure in the event of a LOCA. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The decrease of the minimuin temperature is a passive change to the RHR system. The lower design temperature on ASME code' compliance primarily affects materialimpact testing and the Lowest Service Temperature. Since the lowest service temperature of 30 F is less than the minimum design temperature of 32'F, this change does not increase the probability or consequences of a malfunction'of equipment important to safety previously evaluated in the USAR. There is no change to the system configuration, function, or operation and does not impact any other system, structure,'or component. Therefore, this change does not create the possibility of an accident or I
malfunction of equipment important to safety of a different type. The reduction of the RHR heat

    ..                                                                                    Attachment 2 to U-603276 Page 213 of 252 exchanger shell side minimum design temperature has no affect on system operating parameters.
This activity'does not degrade the reliability or functionality of the RHR heat exchangers.
Therefore, this activity does not reduce'a margin of safety as defined in the basis for any Technical
        . Specification.
BACK-UP AIR SUPPLY TO THE FUEL POOL GATE SEALS
        ' Activity Evaluatedf ECN 31455; USAR Change 8-282                                Log Number: 99-051 Engineering Change Notice (ECN) 31455 installs a back-up air supply for the fuel pool gate seals.
The back-up air station consists of two air bottles along with associated piping, valves, and-controls and is connected to the service air header. Also, this change installs a check valve on the
       '. service air header upstream of the air station to prevent the loss of back-up air in the event of a loss of service air. Updated Safety Analysis Report (USAR) Section 3.6 discusses high-energy and moderate-energy piping failures. If the pressure regulator should fail high, relief valves at the air station protect the piping from overpressurization. Pressure regulators and relief valves are
        . provided at the pneumatic control stations to prevent overpressurization of the fuel pool gate seals. A check valve at the air station prevents the loss of service air header pressure should a relief valve at the air station fail open. If the check valve were to stick open, the system would work the same as it would before this modification. Therefore, this change does not increase the      !
probability or consequences'of an accident previously evaluated in the USAR. This activity
        = primarily affects the service air system which is not considered equipment important to safety.-
However, the fuel pool gate seals are an integral pan of the spent fuel storage pool, which is a stmeture imponant to safety. The addition of a back-up air supply enhances the reliability of service air supply to the fuel pool gate seals by providing an alternate source of makeup air to the seals in the event that normal service air supply pressure is lost. It does not impact the
       ' redundancy of the fuel pool gate seals, the normal or emergency make-up water sources, or the ability of plant operations to restore service air. The addition of the back-up air supply assists in ensuring that USAR Section 9.13.3 safety evaluation statements remain valid. Therefore, this change does 'not increase the probability or consequences of a malfunction of equipment important. to safety previously evaluated in the USAR. Adequate separation has been maintained between new equipment and any other plant system. The air bottles are seismically supported to prevent them from becoming a potential internally generated missile. The back-up air supply does     i not introduce any harmful gasses or chemicals to the environment nor does it create the potential for any additional heat loads.: The original equipment still operates as designed. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type. The Technical Specifications do not discuss the service air system or the fuel pool gate seals. The Technical Specifications do consider fuel pool level, among other considerations, and this modification assists in ensuring that these requirements will be preserved during postulated failure of the service air header. This activity does not change any setpoints,
        ' design limits, or acceptance values. Therefore, this activit'y does not redu'ce a margin of safety as L..      ' defined in the basis for any Technical Specification.                                                ;
L                                                                                                               l 4
l l
l.
L l

Attachment 2          ,
to U-603276 Page 214 of 252 CPS COMPLIANCE WITH REGULATORY GUIDE 1.47 Activity Evaluated: USAR Change 8-277                                           Log Number: 99-052 Regulatory Guide (RG) 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems," addresses status indication ofinoperable safety related systems. This activity revises Updated Safety Analysis Report (USAR) Section 8.3.2.2.2.3 to allow the use of jumpers and temporary bypasses to be installed in operating procedures as well as maintenance
. and testing procedures. RG 1.47 does not provide guidance for the types of procedures which install temporary bypasses and jumpers. Jumpers and temporary bypasses are installed in accordance with plant procedures and comply with the requirements of ANSI 18.7. The same administrative controls and constraints applied to the maintenance and testing procedures will be
. applied tojumpers and lified leads controlled in operating procedures. These requirements will include system restoration such that no permanent change in plant configuration will result.
Configuration changes to structures, systems, or components fall within the hardware change process and a'r e controlled accordingly, thus, there is no isst'e with design, material, or construction standards. This activity does not change any failure modes or system interfaces.
Individual jumpers and lified leads receive their own screening and evaluation, if necessary, in accordance with the requirements of 10CFR50.59. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This change does not allow for any physical changes to be made to the facility as described in the USAR and.there are no changes in system interface.
- Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This activity does not change any setpoints, design limits, acceptance values, or conditions required for operability of components. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
ADDITION OF THE POST ACCIDENT SAMPLING SYSTEM (PASS) SAMPLE CART ALIGNMENT TRACKS Activity Evalu ted: ECN 31396                                                   Log Number: 99-053 Engineering Change Notice (ECN) 31396 installs alignment tracks at the opening into Sample
. Panel IPS02J. The tracks are for the Post ' Accident Sampling System (PASS) grab sample cask cart and lead to a positien under the PASS plunger / needle assembly. This change ensures proper alignment of the PASS sample cask cart and reduces the time required to obtain a grab sample.
- The PASS system is used to assist in the recovery from design basis accidents and provides no safe shutdown or accident mitigating function. The new tracks do not impact any fission product barriers or source terms. They do provide for enhanced efficiency in obtaining PASS grab
' samples. This improved efliciency results in lower potential personnel dose when obtaining samples. Therefore, this change does'not increase the probability of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The new tracks are seismically qualified, consistent with system and stmetural requirements. The tracks constitute passive 4

O                                                                                      Attachment 2
!                                                                                              to U-603276 Page 215 of 252
)
            ; equipment and thereby introduce no potential transients to the PASS. No other components, systems, or structures are affected by the addition of the alignment tracks. Therefore, this change
            . does not increase the probability or consequences of a malfunction of equipment important to L              safety previously evaluated in the USAR. The only postulated failure mode for the cart alignment
            . tracks is that they break loose from their anchorage to the PASS room floor and strike safe shutdown equipment. The tracks are installed to Category I requirements and there is no safety             .
             ~related equipment in the vicinity. Therefore, this change does not create the possibility of an
,           . accident or malfunction of equipment important to safety of a different type. Technical Specification 5.5.3, Post Accident Sampling, requires controls that include training ofpersonnel, procedures for sampling and analysis, and provisions for maintenance of related equipment. The addition of the tracks requires that Clinton Power Station Procedure 6005.01, " Post Accident Sampling," be changed, however, this procedure change does not degrade the adherence to the f requirements set forth in the Technical Specifications. The addition of the tracks does not functionally alter the performance of the PASS system. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
             . DE-ENERGIZING A' ND LOCKING CONTAINMENT /DRYWELL ISOLATION VALVES
:1CC071,L 72,73 AND 74 LActivity Evaluated: ECNs 31409,31410, and 31459;                                  Log Number: 99-054          I USAR Change 8-279; and ORM Change 25-10
                                                                       ~
             - Engineering Change Notices'(ECNs) 31409,31410, and 31459 changes the normal position of Containment /Drywell isolation Valves ICC071, ICC072, ICC073, and ICC074 in response to -
             ' issues raised by Generic Letter 89-10. This change de-energizes the valves and locks the switch              !
in the closed position. A locking device has been added to the' actuator and valve handwheel to
             . prevent manual opening of the valve. Bypass switches have been added at the motor control center (MCC) to bypass alarms and indications for the associated valves to prevent false alarms and indications while the valves are de-energized. The Updated Safety Analysis Report (USAR)                )
does not discuss any accident initiated by the Component Cooling Water (CC) system. The safety function of the valves is to provide a Containment / Drywell Isolation Boundary by being closed in the required time subsequent to the initiation of the Drywell isolation signal. This activity changes the safety function from^ active to passive.= By de-energizing and locking the valves, the modification prevents inadvertent opening and ensures that the valves are normally closed during all modes of operation, thereby performing their safety function. This modification requires an
          .   : Operator to either close the breaker at the MCC or to manually open the locked closed valve.
This results in an additional step to open the valves. When situations warrant, there is adequate Jt ime to open the valves using the MCC switch or manually manipulate the valve. Therefore, this L change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The logic and circuitry of these valves do not interlock with other equipment other than the interlock to close these valves during          i
<.             ~t he initiation of the containment /drywellisolation signal. This activity simply converts the valve's
(                normal position to coincide with its safety position and does not add any new structure, system,

s-Attachment 2             ,
to U-603276 Page 216 of 252
             - or component.1 Any concern that the valves may fail to close is no longer a concern since they become manually-operated, locked-closed valves and are administratively controlled. Therefore,
              .this change does not create the possibility of an accident or malfunction of equipment important
                   ~
              . to safety of a different type than previously evaluated. This change does nat affect the performance or the operability of either the Component Cooling Water System or Containment
             . Isolation System. Technical Specification 3.6.5.3 discusses Drywellisolation. Sealing the valves
             , closed changes the surveillance requirements of the valves, but, the proposed changes are within 1the' scope of the Technical Specification Bases. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REPLACEMENT OF OBSOLETE 480 VAC TELEMECANIQUE CONTACTORS (NEMA SIZES 3 AND 4)                                                                                        -
               ' Activity Evaluatedi ECNs 31517 and 31518                                    Log Number: 99-055 Engineerin'g Change Notices (ECNs) 31517 and 31518 update the design drawings to replace the
         #     Telemecanique A203E48 contactors with a different type of contactor (Cutler / Hammer Model
              . A200).' ' Telemecanique is no longer in business. . The obsolete contactors will be replaced on an as-needed basis.. The'new contactors meet applicable design, material, and construction standards
           ,   applicable to the system and equipment being modified. The contactors are classified and procured as safety related. The activity does not affect the overall system performance in a
: manner which could increase the probability or occurrence of an accident previously evaluated in the Updated Safety Analysis Report (USAR). These contactors are used site-wide in many safety and non-safety related systems Therefore, multiple safety systems may be impacted by this -
change. A' contactor failure interrupts power to affected circuits and equipment, and/or may-propagate the fault to the upstream breakers and circuits. However, the replacement contactors 3              meet the same design standards and function similar to the original contactors. Therefore, this -
change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the US AR. The new contactors meet the same safety L and quality design standards as other components installed in the applicable circuits. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type. The Technical Specifications do not define any margin of safety'for these contactors. Technical Specification 3.8 discusses the requirements for the electrical power systems.; This design change does not alter this Technical Specification requirement; nor, does it      j invalidate or impact the basis for establishing component or system operability Therefore, this
: activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
              ~ MSIV LEAKAGE CONTROL SYSTEM PIPE HEATER WATTAGE CHANGE F            _ Activity Evaluated: ECN 31498; USAR Change 8-294                              Log Number: 99-056 Engineering Change Notice (ECN) 31498 revises plant drawings to show the correct as-built wattage for Main Steam Isolation Valve (MSIV) Leakage Control System (LCS) Pipe Heaters.
The change involves doubling the wattage shown on the drawings, and increasing the loads that

y f     ..
   .-                                                                                       Attachment 2 to U-603276 l                                                                                           Page 217 of 252 L
1 l        . correspond to the heaters in Updated Safety Analysis Report (USAR) Table 8.3-13. The L
wattages are being doubled since the MSIV LCS Pipe Heaters are dual units with two heaters and
: l.         a review of the calculations showed that they only considered a single heater arrangement.
I Appropriate breaker / fuse coordination is provided to prevent a loss of power event or fault
         . propagation event. Failure of the LCS is not an initiator for a Loss of Coolant Accident (LOCA),
instead, the MSIV LCS is used for mitigating a LOCA. The cabling associated 'with the MSIV            -
LCS Pipe Heater has been determined to be adequately sized to accommodate the higher load values. A~ review of the vendor manual, plant drawings, calculations, and Technical Specifications revealed that the dual heater arrangement is the intended, original design, and the design installed in the plant.' These changes are document changes to reflect the as-built configuration of the MSIV LCS Pipe Heaters; there are no physical changes being made to the plant. One of the fission product transport pathways to the environment after a LOCA is leakage through the MSIV
                                                             ~
LCS to the Standby Gas Treatment System. A failure of the Upstream MSIV LCS Pipe Heaters is accommodated by the use of the Downstream MSIV LCS. The Upstream and Downstream MSIV LCS systems are powered from Division 1 and 2 Class IE power systems, respectively.
The change in wattage for the Upstream MSIV LCS Pipe Heater does not adversely affect the failure mode; because should a failure of the Upstream MSIV LCS system occur, proper isolation, redundancy, and separation is maintained between divisions to preclude a failure of one division from inhibitirig the ability of the other divisions from performing their intended functions.
Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This change             j does not add any new stmetures, systems, or components or system interactions that would create any new types of accidents or malfunctions of equipment of a different type. Technical Specification 3.6 l.8 discusses MSIV LCS. . Surveillance Requirement 3.6.1.8.3 states that the upstream pipe heaters meet current draw requirements. The current values provided in the bases for the acceptance criteria for the pipe heaters' operability corresponds to dual heaters. Since investigation has found that the original design for the Upstream MSIV LCS is to have dual Pipe Heaters, there is no affect on the operability of the LCS. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
CONTAINMENT ELECTRICAL PENETRATION PROTECTION COORDINATION CURVES
          ' Activity Evaluate' di USAR Change 8-281                                          Log Number: 99-057 Updated Safety Analysis Report (USAR) Change 8-281 updates the containment electrical penetration protection coordination curves to include all revision levels (a composite curve) for     f molded' case circuit breakers. This USAR change incorporates supporting design analyses, does         !
not physically alter any structure, system, or component, and does not impact any design basis        i accidents discussed in Chapters 6 and 15. A breaker or fuse failure could interrupt power to          j j
affected circuits and equipment, could impact the integrity of electrical penetrations, and/or maj propagate the fault to the upstream breakers and circuits. The breakers and fuses described in this   l USAR change meet the applicable design, material, and constructions standards applicable to the system and equipment being described. This activity does not affect overall safety system I                             '
l l

L:
Attachment 2              ,
to U-603276 Page 218 of 252 performance. Containment is used as a barrier for control of fission material. Integrity of the containment electrical penetrations is maintained by the use of redundant protective devices (breakers and/or fuses). . The protective devices described in this change ensure that the electrical penetrations do not heat-up to the point where a failure of penetration may occur. This USAR
      . change adds the missing breakers and fuses _to USAR Figure 8.3-9 and does not impact the protective function'of the containment electrical penetration protective devices. Therefore, this activity does not increase the probability or consequences of an accident previously evaluated in
      ' the USAR. This activity does not alter the fimetion or operation of the safety systems, and does not der,rade or prevent these systems from performing their intended design function. The fuse failure modes include failure to open, premature opening, and delayed opening. The breakers.and fuses described by this USAR change meet the same design standards as other breakers and fuses shown in Figure 8.3-9, therefore, this activity does not create the possibility of an accident or :
malfunction of equipment important to safety of a different type than previously evaluated. The Technical Specification Bases do not define any margins of safety for the electrical penetrations; therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical
       ' Specification.
ELIMINATION OF USAR LIMITATION TO USE TREMCO DYMERIC PRODUCTS
        - Activity Evaluated:. USAR Change 8-285                                          Log Number: 99-059 This activity revises Updated Safety Analysis Report (USAR) Appendix E, Section 3.1.2.2.10 to
        ' delete the text which limits gap seal materials to Tremco Dymeric products. There are numerous structural gap materials and systems which provide the same fire resistance as the Tremco
   .,    Dymeric detail. This activity allows for a material substitution in the fire rated structural gaps, provided that the materials m_eet the fire resistance criteria identified in American Society for Testing Materials (ASTM) E-119. There is no change to combustible loading or ignition sources.
        - Any material substitution has to meet the testing requirements as specified in the USAR. The design, materials, and construction standards for fire rated gap scalant materials were limited to
       - the Dymeric System as manufactured by Tremco, with the stipulation that the system meets ASTM E-119 fire testing for a 3-hour fire barrier. ' ASTM E-119 Standard Test method for Fire Tests ofBuilding Construction and materials provides the criteria for assembling, testing, documenting and qualifying fire resistant building materials. The alternate materials will provide the same fire protection capability as the Tremco Dymeric, and as such, will not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. . Since any alternate material used must meet the same design and testing requirements, it will provide the same availability, operability, function, and reliability as the Tremco Dymeric materials. Also, these materials must meet the same ASTM E-119 fire testing requirements and provide the same fire resistive capabilities as Tremco Dymeric material.
Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. There are no explicit margins of safety delineated regarding fire rated gap seals or the Fire Protection Program supporting any Technical Specificationc Therefore, this activity.does not reduce a margin of safety as defined in the basis for any Technical Specification.
* Attachment 2 to U403276 Page 219 of 252 INSTALLATION OF A FREEZE SEAL AND TEMPORARY PIPE SUPPORTS DURING                                         ,
THE REPLACEMENT OF CONTAINMENT POOL SUPPLY HEADER ISOLATION VALVE 1FC157
          . Activity Evaluated:= MWO D30172 .                                               Log Number: 99-061 Maintenance Work Order (MWO) D30172 will utilize a freeze seal to provide isolation of the manual Containment Pool Supply Header Isolation Valve 1FC157 due to it leaking by its seat.
         '.During the replacement of this valve, the Fuel Pool Cooling and Clean-up (FPC&C) system will continue to be in service and contingencies wiU be in place for the system. Flow to the upper containment pools will be isolated during the performance of this activity by closing containment
          ' isolation valve' IFC036. The freeze seal will be in place prior to installing temporary supports and removing the valve. Alternate cooling methods to both pools from interconnected systems will be available and are not affected by this activity. This activity will be completed in Mode 4 only.
Installing a freeze seal and temporary supports will allow maintenance activities t'o be performed i on the system without compromising the availability of FPC&C system.
No credible failure modes exist that would initiate an accident or transient previously evaluated in the Updated Safety Analysis Report (USAR). The FPC&C pumps will be stopped by the Main -
Control Room (MCR) operators if the freeze seal should fail. The freeze seal has been evaluated and determined to not degrade the availability, reliability, quality, or aTect overall system performance of the piping following restoraiion. Freeze seal activities will be installed in accordance with applicable plant procedures. The spent fuel pool is designed so that no single failure of stmetures on equipment will cause inability to maintain irradiated fuel submerged in
      ,   ' water or to establish normal fuel pool water level. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of cquipment important to safety previously evaluated in the USAR. If the freeze seal fails, the line would flood the Fuel Building;
          . this failure mode has been' analyzed in USAR Section D3.6.4 for internal flooding due to
         ' moderate energy'line breaks. In operational Mode 4, not all FPC&C system design features are I
         - requiced. Installing a freeze seal will not prevent MCR personnel from monitoring annunciation, indication, and instrumentation that is affected by the' system. The replacement valve is an exact duplicate of the original valve. The system's reliability, performance, or availability is not          j
: impacted by this activity following restoration. Therefore, this activity does not create the           i possibility of an accident or malfunction of equipment important to safety of a different type than
          . previously evaluated in the USAR. This~ activity does not affect any Technical Specification safety     !
limit, limitirg safety system setting, or limiting condition for operation. This activity will not      !
        , ' degrade the ability to satisfy Technical Specification requirement's for the FPC&C system.              ;
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.'                                                                                        )
1

Attachment 2             ,
to U-603276 Page 220 of 252 ORGANIZATIONAL CHANGE ESTABLISHING ASSISTANT P( ANT MANAGER -
OPERATIONS Activity Evaluated: USAR Change 8-290                                          Log Number: 99-062 This Updated Safety Analysis Repon (USAR) change establishes an Assistant Plant Manager-Operations position reponing to the Manager-Clinton Power Station (CPS). The Assistant Plant Manager-Operations will repon to the Manager-CPS, while the Director-Plant Operations will report to the Assistant Plant Manager-Operations. Other changes include:
removing the position of Supervisor-Owner Controlled Area (OCA) Maintenance from USAR Section 13.1.2.1.1.3 and cha.ging the title of Director-Industrial Safety to Supervisor-Industrial Safety. Specific duties and responsibilities of the Manager-CPS and the Director-Plant Operations are unchanged. The Assistant Plant Manager-Operations will assist in the performance ofduties of the Manager-CPS in the area ofplant operations. The Supervisor-OCA Maintenance now reports to the Supervisor-Facilities Group. Since the Supervisor-OCA Maintenance no longer reports directly to the Director-Plant Support Services, this position is no longer required to be referenced in the USAR. This is an administrative organizational change which does not modify plant design or operation. No change is being made to the operation of the facility or to the availability of any equipment. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR, nor does it create the possibility of an accident or malfunction of equipment imponant to safety of a different type. Since this change does not alter the responsibilities of the Plant Manager-CPS, all Technical Specification requirements discussing the dutics, responsibilities, and minimum requirements of the Plant Manager are unchanged.
Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
I~NCORPORATION OE REGULATORY GUIDE 1.149 AND UPDATE ANSI /ANS-3.5 TO A CURRENT REVISION Activity Evaluated: USAR Change 8-292                                          Log Number: 99-067 This change updates Updated Safety Analysis Report (USAR) Section 1.8 to include Regulatory Guide 1.149, Revision 2," Nuclear Power Plant Simulation Facilities For Use In Operating License Examinations," and the Clinton Power Station (CPS) project position for this Regulatory Guide. Also, this change updates USAR Section 13.2.1 to show that ANSI /ANS-3.5,1993,
" Nuclear Power Plant Simulators for Use in Operator Training and Examination," is the revision that the CPS Simulator is being certified to. The document changes affect only the certification criteria of the CPS Simulator and do not have any afTect on acceptance criteria, nor do these changes afTect any actions which are used to mitigate the affects of an accident or transient. This documentation change does not directly affect any plant system, and so, does not impcct the design, material, and construction standards, or afTect the overall system performance of any plant system. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This change O

y .
a
     .                                                                                      Attachment 2         I to U-603276           I Page 221 of 252      j does not introduce any new system, structure, or component (SSC) and does not impact the failure modes or mechanisms for any component or system in the plant. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than previously evaluated. This activity impacts the certification criteria of the CPS Simulator; it does not cause any SSC to operate outside ofits design limits, operational transients in any system, or adverse 1,ystem interaction with any other systems. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
GOULD ITE TYPE HE 600V MOLDED CASE CIRCUlT BREAKER REPLACEMENT Activity Evaluated: ECN 31572; USAR Change 8-295                                   Log Number: 99-070 Engineering Change Notice (ECN) 31572 provides qualified alternate circuit breakers for replacement of the Gould ITE type HE molded case circuit breakers used in 480VAC rated applications. The replacement breakers are Siemens type ED6 and Westinghouse type HFD breakers. The molded case circuit breakers provide power supply continuity and overcurrent protective functions for various 480VAC safety related and non-safety related circuits. The breakers used in safety related applications are relied upon to support mitigation of the accidents     i I
and transients described in Chapters 6 and 15 of the Updated Safety Analysis Report (USAR).
The replacement circuit breakers have been evaluated to be equivalent to or better than the Gould ITE type HE breakers in design, function, operation, and electrical characteristics. To ensure reliability, the breakers were purchased with the appropriate seismic, environmental, technical, and quality requirements. The replacement breakers have equivalent electrical characteristics, including tripping characteristics; thus, there is no impact on the associated circuits, components or systems, and any related response characteristics and interactions. There is no reduction in the    i I
level of protection provided by the replacement breakers to be used as containment penetration overcurrent devices. Because the replacement breakers are functionally equivalent, there are no changes in system performance that would result in operation outside of design limits or contribute to system operational transients. Therefore, this change does not increase the              l probability or consequences of an accident or malfunction of equipment important to safety             i previously evaluated in the USAR. There are no new failure modes introduced as a result of using the replacement breakers; hence, there are no possible methods ofinitiating ary new sequence of events. Additionally, this activity does not introduce any new system interactions.
Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than previously evaluated. A margin of safety defined as
          " sufficient capacity, capability, redundancy, and reliability" is established in the Bases ofImproved l Technical Specification 3.8.1. E!~ trical calculations indicate that suflicient capacity and           ,
l capability is provided by the replacement breakers to perform their safety functions; and the existing margin of protection of cables, loads, breakers, and electrical penetrations is maintained.
Redundancy is not impacted and reliability is maintained by purchasing qualified breakers for safety related applications and performing the required inspections and preventive maintenance         ;
activities. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
S

                                                        ~
Attachment 2
* to U-603276 Page 222 of 252 l
MAINTENANCE ORGANIZATION REPORTING CHANGE y
L Activity Evaluatedi. USAR Change 8-293                                            Log Number: 99-071 This Updated Safety Analysis Report (USAR) establishes the reporting relationship for the Fix-It-Now (FIN) Team to the Manager - Mairitenance. USAR Section 13.1.2.1.3 is being L revised to reflect the transfer of responsibilities for *he FIN Team activities from the Director -
: Electrical Maintenance to Director - FIN Team. This administrative organizational change 'does l L not impact plant design or operatioriof any plant equipment. The core functions and g         responsibilities of the Manager - Maintenance and Director - FIN Team are being maintained.    ~
This USAR change does not affect any structures, systems, or components (SSCs), nor does it
            ' add any SSCs.' Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. Nor, can the change create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR There are no operational
           - considerations associated with the organizational changes being made; therefore, these changes do      ,
           - not reduce a margin of safety as defined in the basis for any Technical Specification.
REVISE USAR TABLE 6.2-47: CHANGE ITEMS IN THRU LINE LEAKAGE CLASS
            ' Activity Evaluatedi USAR Change 8-297                                          ' Log Number: 99-072-This activity revises thirteen items in Updated Safety Analysis Report (USAR) Table 6.2-47, "Thru Line Leakage Class," from " Leak Class 1" to -" Leak Class 2" and moves the test that describes the methodology of calculating secondary containment bypass leakage from USAR Section 6.2.4.3.2.3 to USAR Section 6.2.4.3.4. Also, this activity revises the Operational Requirements Manual, Attachment 4, to designate the same thirteen items as Secondary
: Containment Bypass Paths. USAR Table 6.2-1 lists the maximum allowable containment leak rate (La) of 0.65% of the containment volume in 24 hours at the peak containment pressure, Pa.
It also lists the maximum allowable secondary containment bypass . leak rate of 0.08 of La. -
            , Technical Specification 3.6.1.3.8," Primary Containment Isolation Valves," requires verification
            . that the combined leakage rate for all secondary containment bypass leakage paths is less than or equal 10 0.08 La when pressurized to greater than or equal to Pa. This activity changes the scope of the items which are evaluated in the containment penetration leak test program. They result in an increase in the tabulated secondary containment bypass leak rate. However, the tabulated
            . secondary containment bypass leak rate will be well within the Clinton Power Station licensing basis limit / Continued compliance with the limit for assumed release rates results in no increase in postulated accident consequences. This activity does not nsult in changes in the way that plant equipment and systems are operated. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.L Since this' activity does not introduce any new structure, system, or component, or system interfaces, it does not create the possibility of an accident or malfunction of equipment
             . important to safety of a different type than previously evaluated. The Bases for Technical Specification 3.6.1.3.8 discusses meeting the assumptione in the radiological evaluations, but does

  .                                                                                   Attachment 2 to U-603276 Page 223 of 252 not discuss a margin of safety. Therefore, this activity does not reduce a margin of safety as defmed in the basis for any Technical Specification.
DIVISION 11 DIESEL GENERATOR PARALLEL DROOP COMPENSATION CIRCUlTRY CliANGE Activity Evaluated: ECN 31577; USAR Change 8-307                                Log Number: 99-073 Engineering Change Notice (ECN) 31577 inserts a relay into the parallel droop compensation circuit for the Division 11 Diesel Generator (DG). This eliminates the inherent resistance of the cabling for the auxiliary contacts used to enable or defeat the droop circuit. Additionally, the logic used to defeat or enable the droop circuit changes from normally closed contacts to normally open contacts to provide fail safe protection for the safety function of the DG Most Updated Safety Analysis Report (USAR) Chapter 6 and 15 accidents assume the operation of the DG system for successful accident mitigation; however, no specific USAR accident is initiated by a failure of the DG system operate. The activity meets applicable design, material, and construction standards. The relay is seismically qualified and mounted to assure that it cannot become a missile during seismic events and potentially degrade any other component. The wiring used is the same size and voltage rating, thus eliminating an increase in the potential for fires started for electrical reasons. Post modification testing verified that the modification functioned as anticipated. This activity does not change the fundamental function or objective of the system and does not result in system configurations or operating conditions that could potentially change the DG or other I
systems. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the US AR. The potential failure of the circuit components within the droop circuit are failure of contacts to close when required or failure of the contacts to open when     q 1
required. Regardless of the device driving the contacts, the result of the failures is either shorting across the droop resistor or not shorting across the droop resistor. In either case, the addition of     I the relay does not change the affect of the contact failure upon the voltage regulator. Since the postulated failure modes already exist in the unrevised droop circuitry, this activity does not adversely change, degrade, or prevent actions described in the US AR analysis for mitigating the Lifeci of any transient or accident. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. Since the relays meet the same standards and requirements as the previous relays, there            ,
are no new system interactions, and the effect of the failure is unchanged, this change does not          l create the possibility of an accident or malfunction of equipment important to safety of a different l      type than previously evaluated. This activity does not change any setpoints, design limits, or l      acceptance values discussed in Technical Specifications 3.8.1 and 3.8.2. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.              j I

Attachment 2             ,
to U-603276 Page 224 of 252 y       NON-CONFORMING VOLUMES LISTED IN USAR TABLE 6.2         Activity Evaluated: OE/OD l-98-08-325 and                                       Log Number: 99-074
                                . Tech Spec Bases Change BL-99-011 The net free volumes for the drywell and containment and the suppression pool water volume                  -
available for long term cooling are different than those vclues listed in the Updated Safety
      . Analysis Report (USAR). Operability Evaluation (OE) for 1-98-08 325-OD-1 evaluates and documents the acceptability of these non-conforming conditions. To ensure that the doses, as presented in USAR Section' 15.6.5, to the Main Control Room (MCR) and site bow ries are not exceeded the allowable Main Steam Isolation Valve (MSIV) leakage is being reduced from 28 standard cubic feet per hour (scfh) to 27 scfh. The smaller drywell and containment volumes and the slightly larger suppression pool volume impact the evaluations of the drywell, containment, and suppression pool responses during the postulated accident conditions described in USAR
                       ~
Section 6.2.1. However, the changing volumes are not accident initiators. With the administrative restriction for allowable drywell bypass leakage, only the drywell peak pressure and peak negative pressure are adversely impacted; but the impacts are still below design values.
      - Measured drywell leakage is below the administrative restriction for allowable drywell leakage.
Thus, the drywell and containment will still perform their design basis functions. Other accidents evaluated in USAR Chapter 15 are not impacted, since these accidents are not dependent on the drywell and containment volumes. The environmental qualifications for equipment are maintained and the equipment will still operate as previously evaluated. The administrative restrictions are necessary to ensure that the containment design pressure is not exceeded and to ensure that the doses in the MCR and at site boundaries do not exceed those specified in the USAR. The
      - Emergency Core Cooling System (ECCS) response to a Loss-of-Coolant Accident (LOCA) is not dependent on drywell and containment, thus the ECCS will still provide water to cover and cool the fuel, Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to cafety previously evaluated in the USAR. The drywell and containment are still within their design parameters and no other component or system response is adversely affected. This activity does not introduce any new components or systems into the plant. Therefore, this activity does not create the possibility of an accident or malfunction
       'of equipment important to safety of a different type than previously evaluated. The current                  i allowable MSIV leakage of 28 scfh is specified in Technical Specification Surveillance Requirement 3,6.1.3.9. The administrative restriction to reduce this value to 27 scfh will ensure that the dose to the MCR and site boundaries following a design basis LOCA will remain within the licensing basis. Reducing the allowable leakage to below the current acceptance limits is a conservative measure is in accordance with Nuclear Regulatory Commission guidance. Current, and past test results for both the drywell leakage tests and the MSIV leakage tests are below these          i administrative limits. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
l l

1
:.                                                                                       Attachment 2 to U-603276 Page 225 of 252
        ,   < UPDATE OF DIESEL GENERATOR LOADINGS AND USE OF THE 2000-HOUR DIESEL
            - GENERATOR RATING-Activity Evaluated: USAR Change 8-294
                                     .                                                        Log Number: 99-075
            . Updated Safety' Analysis Report (USAR) Change 8-294 updates the US AR to reflect the                    -
maximum expected coincidental loading of the Division 1, Division 2 and Division 3 Diesel Generators (DGs). 'Also, the USAR is being revised to utilize the 2000-hour rating of the DGs rather than the continuous rating for illustrating the ability of the DGs to handle the maximum coincident loading expected following an accident; This assumption is reasonable, based on its ability to easily exceed an assumed 30 day (720 hour) loss of offsite power with significant margin afterwards for the restoration of offsite power. Most Updated Safety Analysis Report (USAR)
                                   ~
Chapter 6 and 15 accidents require the operation ~of the DG system for successful accident mitigation; The 2000-hour rating provides adequate DG capacity to mitigate the Design Basis Accident and supports the~ proper operation of the required plant equipment. Calculations demonstrate that the ventilation systems for the DG rooms are adequate to ensure that area temperatures in the DG rooms'do not exceed the maximum allowable temperature and diesel fuel l,          - oil requirements are still valid to provide'the required 7 days of operation. The DGs conti'nue to
'io           perform their intended design function to provide an independent, onsite source of emergency -
            - power to the Division 1,2, and 3 Class IE buses.- No radiological effects or radiological release mitigation paths are modified. The postulated consequences of a failure of one DG will be unchanged, since the consequential effect is based on a complete loss of that DG's division of L           : safety related components. ' Therefore, this change does not increase the probability or l            ' consequences of an accident or malfunction of equipment imponant to safety previously evaluated l              in the USARc Since the diesel generators are pan of a plant support / accident mitigating system,        j they have not been assumed to act as an accident initiator; therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This~ activity does not change any setpoints, design limits, or acceptance values discussed in Technical Specifications 3.8.1,3.8.2, and 3.8.3. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
VC SYSTEM, DIVISION 2, SUPPLY AND RETURN FANS,125 VDC CONTROL POWER                                      ]
             . BREAKER TRIP CIRCU!T FUSE REPLACEMENT Activity Evaluated: ECN 31599                                                  Log Number: 99-076 Engineering Change Notice (ECN) 31599 replaces the 35 Amp fuses and associated fuse Pull-Out c           Fuse Blocks in the control power breaker trip circuits for the Control Room Ventilation System
             '(VC) Division 2 Supply and Return Fans with 30 Amp fuses. This provides consistency between
             . the Division I and Division 2.VC Supply and Return Fans and with other 480 volt breaker trip circuits. ! Adequate bicaker/ fuse coordination, in accordance with applicable Updated Safety Analysis Report (USAR) requirements, industry codes and standards, and regulatory guidance, is
: provided upstream of the breakers / fuses to preclude any failures of the VC system from
              . propagating through the Class IE Power System and initiating a Loss of AC Power event. The
                               ,                                                                                        l l
i i

                                                         ~
Anachment 2              ,
i to U-603276 E                                                                                           Page 226 of 252 VC system is required for control room habitability during normal and accidert conditions. The replacement fuses have the same voltage rating and interrupt rating as the previous fuses The
         = 30 Amp fuse opens sooner than the 35 Amp fuse, thus the cabling and. equipment will be better -
         - protected. The reduction of the fuse rating does not adversely affect the protection capabilities of
          ' the fuses. The lower amperage rating has been demonstrated as acceptable for its circuit application.; The replacement fuses will maintain the appropriate separation, independence, and
          ' isolation. The replacement fuses and their associated fuse blocks are of a type already utilized in the plantJ The VC system will operate as designed to maintain the habitability of the control room after Design Basis Accidentsc The VC system is not required to mitigate any radiological consequences outside of control room habitability. Therefore, this change does not increase the
: probability or' consequences of an accident or malfunction of equipment important to safety
         - previously eval _uated in the USAR. No new system interactions, or new failure modes are introduced by this change.2 Therefore, this change does not create the possibility of an accident or malfunction of equipment important to_ safety of a different type than previously evaluated.
Technical Specifications 3.3,7.1 and 3.7.3 discuss " Control Room Ventilation System Instrumentation" and " Control Room Ventilation System," respectively, There are no design
          - limitations provided in these Specifications that apply to the changes being made to the VC system., The changes do not affect any safety limits, limiting safety system settings, limiting
         - conditions for operation, or surveillance requirements applicable to the VC system. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical 4
Specification.
                                 ~
ANSI N509I510 VARIANCE REPORT
           = Activity Evaluated: HVAC-01-CP, Rev. 0; USAR Change 8-304                      Log Number: 99-077 This activity involves a clarification to the requirements in ANSI Standard N509-1976 as interpreted and applied by Clinton Power Station (CPS). The variance report, which is ieferenced in Updated Safety Analysis Report (USAR) Section 1.8, lists exceptions to the requirements of ANSI N509/510 at CPS. This activity removes some editorial errors, enhances and clarifies
            . vague or deficient engineeringjustification for the design, manufacture, testing and acceptance criteria variances from ANSI N509,'_' Air Cleaning Units and Components," and ANSI N510,
            " Testing of Nuclear Air Cleaning Systems," standards. Both the Standby Gas Treatment System (SGTS) and the Main Control Room (MCR) Heating Ventilation and Air Conditioning (HVAC)
System (VC) filtiation units are designed to mitigate the consequences of postulated accidents.
Failure of these systems cannot act as an initiating event for any of the_ design basis accidents identified in USAR Chapter 6 or 15 This activity does not change the design, operation or testing of SGTS, VC, and other atmospheric filter units. Thus, this activity does not affect the overall performance of the atmospheric filter units nor cause the systems to operate outside their des @s limits? This activity corrects editorial errors and clarifies the engineering justification for some of the deviations from Regulatory Guide 1.52. It does not alter the_ design, testing, or L operation of equipment important to safety. Therefore, this change does not increase the probability or consequences'of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The editorial errors in the variance report and the deficient

      , o. ,
Attachment 2 to U-603276 Page 227 of 252 engineering justification to the variance therein have no direct or indirect effects on the ability of      1 the SGTS and VC system to perform their safety function. Correction of the editorial error and the enhancement of the engineeringjustification to the variance,do not change the design, testing, and operation of either the SGTS or VC system. This activity does not introduce any component or system into the plant, nor change any component or system already in the plant. Therefore, this activity does not create the possibility of an accident or malfunction of equbment important         .
to safety of a different type than previously evaluated. This activity does not directly or indirectly
               ; affect any acceptance values, safety limits, limiting safety system settings, limiting conditions for operation, or design limitations associated wiscomponents involved in this activity. Therefore, this activity does not reduce a margin of safety as dermed in the basis for any Technical Specification
                . USAR CHANGE REGARDING MSIV HIGH STEAM LINE TEMPERATURE CLOSURE                                             {
Activity Evaluated: USAR Change 8-298 -                                            Log Number: 99-078 1
This activity evaluates the combined effect of two Engineering Change Notices (ECNs) (29926 and 30032). The combined effect is that the Main Steam Isolation Valves (MSIVs) no longer go closed on a spurious High Steam Line Temperature signal following a Loss of AC Power. As a                 j
                . result, the Updated Safety Analysis Report (USAR) Chapter 15 ar,alysis for these events is impacted.f Since the leak detection system will not generate a trip signal following a Loss of AC Power, the MSIVs will not start to close immediately after the power loss. This impacts the -
sequence of events following a loss of AC power, however, this does not alter the probability of occurrence. ECNs 29926 and 30032 added the Main Steam Line High Temperature Modules as new loads on the Division 1 and 2 Nuclear System Power System (NSPS) buses and removed
                 ' them as loads from the Divisional buses. The impact of these load changes were evaluated in Safety Evaluations 99-011 and 99-026. There is no additional impact on equipment important to l                  safety due to the combined effect of these ECNs. This change allows the MSIVs to stay open L                  during interruptions of AC power; however, the MSIVs do not need to serve as a radiological
                                       ~
L                  barrier during this time for this event. MSIV closure to limit radiological releases is initiated by
                ' methods other than through loss of power to the leak detection system logic. These methods of isolating the MSIVs to prevent a radiological release are not affected by this change. Therefore, this activity does not increase the probability'or consequences of an accident or malfunction of
                  . equipment important to safety previously' evaluated in the USAR. No new equipment is being added by this activity, and'no equipment is being modified. The t.nly impact is the removal of the immediate closure time associated with the MSIVs on a Loss of AC Power. MSIV closure is
                 . expected to be nominally at t.= 28 seconds due to a loss of the main condenser vacuum, rather than t = 0 seconds per current design.. The USAR evaluates inadvertent MSIV closuresi In additica, a search of the licensing files has ' revealed that the original design basis for the loss of
                   'AC power transient did not account for an immediate closure of the MSIVs. It was change                   j package 216 in Revision 0 of the USAR that the design basis was altered to incorporate the possibility that the MSIV closure may be initiated immediately on a loss of AC power transient due to the spurious leak detection system trip. By making the current changes, which climinates the spurious leak detection system trip, CPS will be reming to the original design basis. A L.

Aitachment 2              ,
to U-603276 Page 228 of 252 '
search of the CPS licensing files has shown that there were no other changes made to this design
; basis since 1988, and hence a~ return to the pre-1988 design basis is possible without evaluating its impact on the subsequent design changes. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR, This activity does not directly or indirectly affect any acceptance values,               .
safety limit's, limiting safety system settings, limiting conditions for operation, or design limitations     -
l
: associated with components involved in this activity. Theretvie, this activity does not reduce a
: margin of safety as defined in the basis for any Tecnical Specification.
4 I
DIESEL GENERATOR (DG) VOLTAGE REGULATOR AND GOVERNOR TEST Astivity byaluatediCPS Procedure 2808.01'                                         Log Number: 99-079 L This activity addresses special test procedure, 2808.01 Revision 0, " Diesel Generator (DG)
. Voltage Regulator and Governor Test," which monitors generator frequency, voltage and enghe speed under various loading conditions to determine the limits of the automatic resets for the voltage regulators and governors, and the voltage regulator response to increasing and decreasing ioads.LInstallation of the test equipment for performance of the special test requires lifting wires for connecting monitoring equipment in parallel with existing circuitry. Data obtained during the test gives the proper assurance that the desired changes in limits of the automatic resets for the voltage regulator and governors and voltage limits are maintained while in the isochronous Mode.
Most Updated Safety Analysis Report (USAR) Chapter 6 and 15 accidents assume the operation of the DG system for successful accident mitigation. This special test essentially represents a DG start, parallel to the bus, and load similar to the DG operation for routine surveillance testing.
The differences that exist are primarily associated with the data collected during the performance of the test. Selecti..g specific components to place on the DG for loading purposes is no different 7
than the normal system operation. Critical parameters are monitored to ensure equipment is not degraded 'during these tests. The test equipment is not IE qualified or seismically qualified, but is temporary and under positive control. The special test imposes appropriate load limits on the DG
, and flow limits on the systems. The special test is performed as an infrequently performed evolutiong which provides for the performance of pre-test briefs and the assignment of extra operating stafras required. Safety related components such as the decay heat removal system are aligned to non-impacted divisions during the conduct of the tests. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.1 The DG system is an accident mitigating system. No                   t credible failure modes that can serve as an initiating event for an accident or malfunction of                 i equipment important to safety are created due to the performance of this special test. No acceptance v' alues, safety limits, limiting safety system settings, limiting conditions for operation, or design limitations associated _with components involved in this activity are negatively impacted.           ;
In general, this test is used to demonstrate compliance with Technical Specification Surveillance Requirementsc Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.-
                                                                                                                    )
1 i
i
                                                                                                                   )

b
     +
i
: i.   ,
Attachment 2 to U-603276 Page 229 of 252 DIVISION I DIESEL GENERATOR PARALLEL DROOP COMPENSATION CIRCUITRY CHANGE l
Activity Evaluated: ECN 31580; USAR Change 8-316                                Log Number: 99-080 Engineering Change Notice (ECN) 31580 inse :s a relay into the parallel droop compensation circuit for the Division I Diesel Generator (DG). This eliminates the inherent resistance of the cabling for the auxiliary contacts used to enable or defeat the droop circuit. Additionally, the logic used to defeat or enable the droop circuit changes from normally closed contacts to normally open contacts to provide fail safe protection for the safety function of the DG. Most Updated Safety Analysis Report (USAR) Chapter 6 and 15 accidents assume the operation of the DG system for successful accident mitigation; however, no specific USAR accident is initiated by a failure of the DG system operate. The activity meets applicable design, material, and construction standards. The relay is seismically qualified and mounted to assure that it cannot become a missile during seismic events and potentially degrade any other compom nt. The wiring used is the same size and voltage rating, thus eliminating an increase in the potential for fires staned for electrical reasons. Post modification testing verified that the modification functioned as anticipated. This activity does not change the fundamental function or objective of the system and does not result in system configurations or operating conditions that could potentially change the DG or other systems. Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR. The potential failure of the circuit components within the droop circuit are failure of contacts to close when required or failure of the contacts to open when required. Regardless of the device driving the contacts, the result of the failures is either shorting across the droop resistor or not shorting across the droop resistor. In either case, the addition of the relay does not change the affect of the contact failure upon the voltage regulator. Since the postulated failure modes already exist in the unrevised droop circuitry, this activity does not adversely change, degrade, or prevent actions described in the USAR analysis for mitigating the afTect of any transient or accident. Therefore, this change does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. Since the relays meet the same standards and requirements as the previous relays, there are no new system interactions, and the effect of the failure is unchanged, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. This activity does not change any setpoints, design limits, or acceptance values discussed in Technical Specifications 3.8.1 and 3.8.2. Therefore, tl-is activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
i INSTRUMENT SETPOINT- AND UNCERTAINTY CR OPERABILITY DETERMINATION AND EVALUATION Actisity' Evaluated: Operability Evaluation 1-98-07-308-OE-2                    Log Number: 99-081 Operability Evaluation (OE) 1-98-07-308-OE-2 confirmed that, in spite of discovery of a significant number of errors and inconsistencies, the existing plant setpoints and indications are still adequately established for all plant operating modes. The scope of this activity includes l

Attachment 2           .
to U-603276 Page 230 of 252 safety-related instrumentation required by Improved Technical Specifications, Emergency -
Operating Procedures l and Regulatory Guides for all plant operating modes. Since the scope of the review' includes the majority of the safety related instrument setpoints and indications, all'
: Updated Safety Analysis Repart (USAR) Chapter 6 and 15 accidents are potentially affected.
This activity has determined that setpoints and instrument indication are adequately adjusted.
These instruments were evaluated and found to be properly adjusted in the field.                          -
The parameters evaluated included instruinent setpoint functions and instrument indications function have two credible failure inodesi setting / indication less conservative than design and setting / indication more conservative than design. The indication functions are not accident initiators.i They are used for operator indications and surveillance procedure acceptance devices.
Similarly, instrument setpoints adjusted less conservatively than design are not considered to be accident: initiators.: An abnormal event or accident would be required to establish plant conditions resulting in actuation of an instrument adjusted to a non-conservative setpoint value. The instrument setpoints providing automatic equipment initiation functions were determined to be adequately established for plant operation by the OE. The original setpoint determinations have not resulted in a history of spurious equipment actuations, and the errors found by the OE did not result in any setpoints exceeding allowable values or approaching analytical limits. Therefore, the
       ' setpoints validated or established by this activity will not change, degrade or prevent actions -
assumed in the accident analysis. The existing plant surveillance procedures and calibration and testing requirements are not altered by this activity, therefore, this activity does not impose more severe or increased testing requirements. In addition, this activity does not reduce system redundancy or independence. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. No setpoints or indications were determined to exceed the applicable Technical
        . Specification allowable values or approach any Technical Specification analytical limits as a result of the inconsistencies determined by the OE. Also, this activity does not introduce any new structures, systems,'o'r components into the plant. Therefore, this activity does not create the possibility of an accident or malfunction o' f equipment imponant to safety of a different type than previously evaluated l The margin of safety, as defined in the context of this evaluation, is considered to be the difference between the analytical limits and the design failure point for the related equipment, system or function. Each safety-related instrumentation function or monitored parameter required to be op.erable was evaluated. The OE determined whether the specified value or setpoint was a nominal value or based on an analytical limit. None of the reviews documented in 1-93-07-308-OE-2 determined an existing plant setpoint that exceeded an allowable value or
       . approached an analytical limit as defined by the Technical Specifications. Therefore, this activity
       . does not reduce a margin of safety as definnt in the basis for any Technical Specification.
FEEDWATER FATlGUE CYCLES OPERABILITY EVALUATION Activity Evaluated: 148-03-495-OE-1                                              Log Number: 99-082
       . Clinton Power. Station Condition Report 1-98-03-495 identified that ASME Code Class 1
: Feedwater (FW) piping may have exceeded the allowable cycles under the Design Fatigue i

        ,c                                                                                            ~ Attachment 2 to U-603276 '
Page 231 of 252 Analysis. The Reactor Pressure Vessel (RPV) FW nozzles were designed for an overall 111 plant shutdowns, which allows one postulated on/off FW injection during the " Hot Standby" and five
: on/offinjection cycles during the " Shutdown" event for each plant shutdown. However, the
      ,            ASME Code Class 1 FW piping design fatigue calculations did not include these same FW                           ,
injection loading assumptions. Preliminary estimates of the number of FW injections experienced by the plant'to date indicate that 436 of these injections have occurred. This Safety Evaluation            -
                  ' demonstrates the operability of the FW system despite having experienced more on/offinjection                i
                  . cycles than originally designed and supports the Operability Determination (OD) and Operability Evaluation (OE) which allows a disposition of"Use As-Is" for the Feedwater system.
This activity does not . impact or alter the FW system configuration, operation, response characteristics, or result in additional operational transients. Calculation EMD-033201 indicates           '
that the Class I components in question will remain operabie for approximately two additional fuel cycles.' Since the Class I components remain within the Cumulative Usage Factor of the                 4
                   'ASME Code and Cumulative Usage Factor of 0.1, this activity does not increase the probability
                 - or consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the Updated Safety Analysis Report (USAR).- This change does not introduce any i
l " new failure modes which could initiate any accident or transient events not previously evaluated in the USAR. This activity does not delete, add, modify, or relocate any equipraent imponant to
                 - safety. Nor does this activity. alter the operational activities or co.ntrols associated with the FW systemaTherefore, this activity does not create the possibility of an accident or malfunction of
: equipment important to safety of a different type than previously evaluated. This activity does not impact any flow biased function of the Reactor Protection System or any thermal limits via changes to FW temperature. No Limiting Conditions for Operation are identified that could be affected by this change. This' activity does not alter the system configuration, response
                  . characteristics, or result in any additional operational transients which could result in a reduction of a margin of safety'as defined in the basis for any Technical Specification.
                 ' DIVISION 111 DIESEL GENERATOR PARALLEL DROOP COMPENSATION CIRCUITRY CHANGE L Activity EvaluatedLECN 30353                                                        Log Number: 99-083
                 . Engineering Change Notice (ECN) 30353 inserts a relay into the parallel droop compensation                    _
circuit for the Division Ill Diesel Generator (DG). This eliminates the inherent resistance of the cabling for the auxiliary contacts used to enable or defeat the droop circuit. Additionally, the          j
                  > logic used to defeat or enable the droop circuit changes from normally closed contacts to normally open contacts to provide fail safe protection for the safety function of the DG.' Most Updated Safety Analysis Report (USAR) Chapter 6 and 15 accidents assume the operation of the DG system for successful accident mitigation; however, no specific USAR accident is initiated by a
( $,_     " failure of the DG system operate. The activit'y meets applicable design, material, and construction W
                                                                                      ~
                  - standards. The relay is seismically qualified and mounted to assure that it cannot become a missile
                  - during seis'mic events and potentially degrade any other component. The wiring used is the same size and voltage rating, thus eliminating an increase in the potential for fires started for electrical     j p
l
                                    \

b y                                                                             .
Attachment 2            ,
to U-603276 Page 232 of 252 reasons. Post modification testing verified that the modification functioned as anticipated. This
           ' activity does not change the fundamental function or objective of the system and does not result in 2
system configurations or operating conditions that could potentially change the DG or other -              i systems 2 Therefore, this change does not increase the probability or consequences of an accident previously evaluated in the USAR.;The potential failure of the circuit components within the droop circuit are failure of contacts to close when requied or failure of the contacts to open when required. Regardless of the device driving the contacts, the result of the failures is either shorting
: across the droop resistor or not shorting across the droop resistor. In either case, the addition of L the relay does not change the affect of the contact failure upon the voltage regulator. Since the postulated failure modes already exist in the unrevised droop circuitry, this ~ activity does not -
adversely change, degrade, or prevent actions described in the USAR analysis for mitigating the affect of.any transient or accident. Therefore, this change does not increase the probability or Econsequences of a malfunction of equipment important to safety previously evaluated in the '
USAR.t$ince the relays meet the same standards and requirements as the previous relays, there are no new system interactions, and the effect of the failure is unchanged, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated This activity does not change any setpoints, design limits, or
        . acceptance values discussed in Technical Specifications 3.8.1 and 3.8.2. Therefore, this activity -
: does not reduce a margin of safety as defined in the basis for any Technical Specification.
          . NON-CONSERVATIVE ALLOWABLE VALUE FOR SAFETY RELATED INSTRUMENT-
          ' SETPOINTS '
           'Activit) Evaluatedi Compensatory ' Actions for CR l-99-03-081                  .. Log Number: 99-084 and OE l-99-03-081-OE-1
: Condition Report (CR) 1-99-03-081 documents discovery that the allowable value (AV) for the
          . Average Power Range Monitors (APRMs) a Fixed Neutron Flux - High Function, as defined in Technical Specification Table 3.3.1.1-1 Item 2c, is not conservative enough to confidently avoid
           .the analytical limit (AL) using current setpoint methodology. Operability Evaluation (OE) 99-03-081-OE-1 was subsequently prepared, which determined that compensatory actions were .
          ' necessary to implement conservative setpoints.' The compensatory action involves restricting the 3 allowable value from s 120% rated thermal power (RTP) to a more conservative value of s 119%
RTP.? No change to the AL is required by this activity. The credible failure mode is considered
: the as-left setpoint adjustment being more conservative than design, such that the new AV may be 1
          . challenged when taking as'-found data during subsequent calibration. The instrument setpoint functions in the scope of the subject activity aie designed to actuate mitigating equipment should an accidedor transient occur, Restricting adjustment of the setpoints within the more e           conservative allowable value presents the potential for an inadvertent actuation of mitigating equipment. l However; the revised AV is based on a 95% error probability limit at a 95%
  ,       . confidence value (as accepted by Regulatory Guide 1.105, Revision 2); thus, ensuring that the uncertainties will not cause a challenge to the analytical limit. This activity provides additional assurance that the equipment operability and system availability assumptions utilized in plant accident analyses will not be impacted. This activity does not degrade design basis performance
{

  .                                                                                       Attachment 2 to U-603276 Page 233 of 252 I
by affecting any environmental, seismic or separation design bases or criteria. This activity does   l not physically or operationally affect any system, structure, or component. Therefore, this activity does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). This activity does not affect
        - overall system performance, change system response characteristics, cause system operation outside of design limits, cause system operational transients, or cause adverse system interactions. ;
No plant equipment was replaced, removed or added as a result of this activity. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The margin of safety, as defined  t in the context of this evaluation, is considered to be the difference between the analytical limits and the design failure point for the related equipment, system or function. Since the subject activity provides an additional conservatism between the analytical limit and the Technical         4 Specification allowable value, the margin of safety is not reduced. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.              {
REPLACEMENT SIEMENS TYPE ED6 BREAKERS                                                               i I
,         Activity Evaluated: ECNs 31625 & 31605, and                                      Log Number: 99-087 USAR Change 8-302 1
Engineering Change Notice (ECN) 31605 replaces both of the breakers within breaker cubicle 1 AP75E2C and ECN 31625 replaces both of the breakers within breaker cubicle 1 AP75E3 A.
These cubicles contain two Gould ITE Type HE breakers, with one of the two breakers having a       1 DC shunt trip. Cubicle 1 AP75E2C contains two 50 amp breakers, and cubicle 1 AP75E3A L          contains two 90 amp breakers. These ECNs replace the Gould ITE Type HE breakers with j          Siemens Type ED6 Revision 5 clrcuit breakers. Safety Evaluation 99-070 addresses the use of l          Siemens Type ED6 Revision 5 circuit breakers as qualified replacements for Gould ITE Type HE circuit breakers for 480VAC applications when used in Gould ITE AC Motor Control Centers and ECN 31572 contains a duailed comparison of the electrical characteristics between the breakers. The scope of this activity is limited to the facility changes not addressed by ECN 31572. Thus, the scope is limited to the affects of the wiring changes required to install the breakers; and the relatively small increase in the 125 VDC battery (1DC02E) loading due to the
         - new shunt trip device.                                                                             l No specific Updated Safety Analysis Report (USAR) Chapter 6 or 15 accident is initiated by a failure of the IDC02E battery to operate. The wiring changes are only made to account for the lack of pigtails on the new breakers, and do not represent functional or logic changes to the breaker control circuit. Divisional separation is preserved and the new wiring meets or exceeds the size, voltage rating, and qualification required for this application. This activity does not change the fundamental function or objective of the system and does not result in system configurations or operating conditions that could potentially change the DC or other systems.
Calculations demonstrate that the total battery loads do not exceed the capacity rating of the battery. This activity does not change assumed radiological release limits or fission product barrier performance during postulated accidents. Therefore, this change does not increase the

Attachment 2               ,
to U-603276 Page 234 of 252
       - probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.1 This actid does not create any new system interactions or
       ' alter performance characteristics of the existing system. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. No acceptance values, safety limits, limiting safety system settings, or limiting conditions for operation associated with components are adversely affected by this .
activity, and no design limitations are negatively impacted. Therefore, this activity does not reduce a margin 'of safety as defined in the basis for any Technical Specification.
       - NON-CONSERVATIVE ALLOWABLE VALUES FOR SAFETY RELATED INSTRUMENT
:SETPOINTS.
       . Activity Evaluated: 3199-04-089-OE-1                                            Log Number: 99-088 l Condition Report l_-99-04-089 identified non-conservative allowable values for four instrument setpoint functions:: Reactor Vessel Level 1 Low-Low-Low, Reactor Vessel Level 2 Low-Low, Reactor Vessel Level 3 Low, and Main Steam Line Flow-High. Operability Evaluation (OE) 1-99-04-089-OE-1 addresses the impact ofimplementing the compensatory actions that were
  .      necessary to support plant restart commitments. The compensatory actions include establishing conservative allowable values in the Technical Specifications and revising data sheets for calibration and channel functional procedures. Review of the nominal trip setpoints appearing in the Operational Requirements Manual reveal the existing setpoints are conservative with respect
       - to the newly determined acceptance limits, and thus do not require revision. No field adjustment
        - or re-calibration is necessary to implement the newly determined allowable valus The setpoint a       ~ values themselves are not assumed initiators of any evaluated accident. Therefore, the compensatory actions are not event initiators.1 The associated instrument setpoint functions to mitigate equipment should an accident or transient occur. The plant acceptance criteria being Limplemented for the affected instrument setpoints is based on a 95% error probability limit at a 95% confidence value, or a "2-sigma" value. Regulatory Guide L105, Revision 2, states the Nuclear Regulatory Commission has accepted the use of this probability limit for instrument
       . uncertainty errors. Implementing these compensatory actions ensures that Clinton Power Station will be bounded by the acceptance criterion. Since the subject activity revises allowable values in
        .the conservative direction, without affecting the trip setpoints, the attendant safety systems will
                 ~
not be affected for a postulated accident scenario. No design, material and construction standards were impacted by this activity This activity does not affect any environmental, seismic, or
        < separation criteriaJ Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equip' ment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). No plant equipment was replaced, removed or added as a result of this activity. This activity does not change the function or method of operation of any plant equipment Therefore, this activity does not create the possibility of an accident or malfunction of
       . equipment importe.nt to safety of a different type than previously evaluated in the USAR. The margin of safety as defined in the context of this evaluation is considered to be the difference between the analytical limita and the design failure point for the related equipment, system, or function.L Those parameters with analytical limits', which are primarily limited to the automatically F4

W,
: c.                                ,
     +
Attachment 2 to U-603276 Page 235 of 252 initiated functions, have this information documented in the appropriate Technical Specification and Technical Specification Bases sectionsc Since the activity provides an additional conservatism
              . between the analytical limit and the Technical Specification allowable value, the margin of safety L             . is not adversely impacted. lTherefore, this activity does not reduce a margin of safety as defmed in
              ~ the basis for any Technical Specification.
               ' NON-CONSERVATIVE ALLOWABLE VALUES FOR SAFETY RELATED INSTRUMENT 7        LSETPOINTS                                                                                                      '
!-            -- Activity Evaluated:' Compensatory Actions for CR l-99-03-361                      Log Number: 99-089 and OE l'-99-03-361-OE-1; and ORM Change 25-13                                                               ,
Condition Report (CR) 1 99-03-361 documents discovery that the allowable values (AV) for                    f
: three instrument setpoint functions are less than current method 'ogy permits: Safety Relief '
L              . Valve (SRV) Low-Low Set, Low Pressure Core Spray (LPCS) Pump discharge Pressure - High, L                and Low Pressure Coolant Injection (LPCI) Pump Dis-harge Pressure - High. Operability Evaluation (OE) 1-99-03-361-OE-1 was subsequently prepared, which determined that compensatory' actions were necessary to implement conservative setpoints. The compensatory actions include revising the affected AV and~ nominal trip setpoints in the Technical Specifications, Operational Requirements Manual, and data sheets for calibration procedures. This activity does not change any setpoint already established in the' field and thus does~ not change any credible u               failure mode. The instrument setpoint functions in the scope of the subject activity are designed l               to actuats mitigating equipment should an accident or transient occur The revised AV is based I~
on a 95% error probability limit at a 95% confidence value (as accepted by Regulatory Guide 1.105 Revision 2); thus, ensuring that the uncertainties will not cause 'a challenge to the analytical limit.' . atis activity provides additional assurance that the equipment operability and system L                availability assumptions utilized in plant accident analyses will not be impacted. This activity does
?             - not degrade design basis performance by affecting any environmental, seismic or separation design L                bises or criteria. This activity does not physically _ or operationally affect any system, structure, or l                component. Therefore, this activity does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis h               Report (USAR). <This activity does not affect overall system performance, change system                      j H
response characteristics, cause system operation outside ofdesign limits, cause system operational         i transients, or cause adverse system interactions. No plant equiprhent 'was replaced, removed or            I
,             - added as~ a result of this activity. ' Therefore, this activity' does not create the possibility of an L             . accident'or malfunction of equipment important to safety of a different type than previously                 4 evaluated in the USAR.1 The margin of safety, as definbd in the context of this evaluation, is
!             s considered to be the difference between the analytical limits and the~ design failure point for the
                                                ~
related equipment, system or function.-' Since the subject activity provides an additional conservatism between the analytical limit'and the Technical Specification allowable value, the
                . margin of safety is not reduced. Therefore, this activity does not reduce a margin of safety as p ,'             defined 'in the basis for any Technical Specification. ,
L

Attachment 2             ,
to U-603276 Page 236 of 252 REMOVAL OF DIVISION 2125VDC BATTERY CHARGER 1B GROUND FAULT.
       ' DETECTION LIGHTS Ac;ivity Evaluated: Temporary Modification 99-034                             Log Number: 99-090-3 Temporary Modification 99-034 removes both Ground Fault Indicating Lights from the Division 2125VDC Battery Charger IB (IDC07E). The removal of the indicating lights will improve the sensitivity of the Ground Detecting Voltmeter and Recorder installed in the Division
        -2125VDC Motor Control Center (MCC) IB. This Temporary Modification removes the light bulbs only and leaves the wiring and sockets in place. . Section 8.3.2.1.2.4 of the Updated Safety
: Analysis Report (USAR) provides a listing and brief functional description of the instrumentation which is installed in 125VDC MCC 1B to provide sufficient monitoring of the DC System status
        'and its readiness to perform its cited function. Not mentioned in the USAR, but provided as a feature of the Battery Charger, are 2 indicating lights which provide a non-audible local indication of a ground fault on the bus. Loss orlocal ground fault indication at the battery charger is not
       . discussed in either Chapter 6 or 15 of the US AR. Removal of the indicating lights will not idegrade the overall ground detecting circuit since these lamps are only provided by the manufacturer as an enhancement for charger operation. Therefore, this activity does not increase
: the probability or consequences of an accident or malfunction of equipment important to safety
       ' previously evaluated in the.USAR. The impact of bulb removal is the same as a burned out bulb in the circuit No new failure modes or system interactions are being introduced. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously_ evaluated." This change does not degrade the basis for operability requirements established for the 125 VDC electrical distribution system in Technical
       ' Specifications. The Technical Specifications do not provide any basis or requirements for the DC system ground detection at the Battery Charger. Therefore, this activity does not reduce a margin
: of safety as defined in the basis for any Technical Specification.
NON-CONSERVATIVE ALLOWABLE VALUES FOR INSTRUMENT SETPOINTS FOR THE ATWS REACTOR VESSEL LEVEL 2 LOW-LOW FUNCTION Activity. Evaluated:- Compensatory Actions for CR l-99-04-173                 Log Number: 99-091-and OE l-99-04-173-OE-1
       - Condition Report (CR) 1-99-04 '173 documents discovery of a non-conservative allowable value (AV) for the_ Anticipated Transient _Without Scram (ATWS)- Reactor Vessel Level 2 Low-Low
       ~ instmment setpoint function ~. Operability Evaluation (OE) 1-99-04-173-OE-1 was subsequently             ,
prepared,1which determined that compensatory measures were necessary to support plant restart            i
        ; commitments. The compensatory actions include revising the affected AV to 2 49 inches in the Technical Specifications and data sheets for Clinton Power Station calibration procedure 9434.01.        .
This activity does not change any setpoint already established in the field and, thus, does not          l change any credible failure mode. The instrument setpoint functions in the scope of the subject          ;
activity are designed to actuate mitigating equipment should an accident or transient occur. The         {
revised AV is based on a 95% error probability limit at a 95% confidence value (as accepted by

7                                                                                      Attachment 2 to U-603276 Page 237 of 252 Regulatory Guide 1.105,' Revision 2). This activity provides additional assurance that the equipment' operability and system availability assumptions utilized in plant accident analyses will not be impacted.' This activity does not degrade design basis performance by affecting any :
environmental, seismic or separation design bases or criteria. Therefore, this activity does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the Updated' Safety Analysis Report (US AR). This activity does not ' affect overall     .
system performance, change system response characteristics, cause system operation outside of :
          - design limits, cause system operational transients, or cause adverse system interactions. No plant equipment was replaced, removed or added as a result of this activity. Therefore, this activity, does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR; The margin of safety, as defined in the context of this evaluation, is considered to be the difference between the analytical limits and the design failure point for the related equipment, system or function. Since the subject activity provides an additional conservatism between the analytical limit and the Technical Specification allowable value, the margin of safety is not impacted.; Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
I
           ' OPERABILITY DETERMINATION 1-99-01-162 AND' COMPENSATORY ACTIONS REQUIRED FOR DEFICIENT KIRK-KEY INTERLOCK Activity Evaluated: 1-99-01-162-OD-1                                           Log Number: 99-092      l l Operability Determination (OD) 1-99-01-162 established compensatory actions and this Safety Evaluation _ evaluates compensatory actions being taken due to a deficient disconnect switch (ET4) in the Auxiliary Power (AP) system which is fed from the Emergency Reserve Auxiliary   .
Transformer (ERAT). The 1 APET4 disconnect switch is interlocked with 1 APET14 which provides isolation for the Reserve Auxiliary Transformer (RAT). The manual" Kirk" key interlo'ck function of I APET4 switch is broken, it will not be permanently deleted, but accepted in
           . its present state with compensatory actions in place until the deficiency is corrected. The following compensatory actions are being implemented in place of the failed manual " Kirk" key interlock to maint'ain or decrease the probability oflosing off-site power:. Procedure 3501.01 will
;            be changed to warn operators that the mechanical" Kirk" key interlock has failed, a tagout will be
           . put in place along with danger and caution tags hung at I APET4 and 1 APET14 to inform operators of degraded conditions, and two padlock keys will be raised to the next administrative control level.
            .This system and the components involved are classified as non-safety related, non-Class lE, and f
L            non-seismic. There are no design basis accidents that could be credibly impacted by this activity.
Compensatory actions being implemented in place of an interlock do not make hardware changes           i or modifications to plant equipment, do not change setpoints to instrumentation, do not affect system performance of any Structure, System, or Component (SSC), and do not place any system in a lineup that has not already been analyzed. All design parameters of t1 ERAT system o       f mentioned in the Updated Safety Analysis Report (USAR) will not change The USAR requires a              l
  ^         . safety or non-safety related systems to be designed so that any failure of a SSC will not cause the

I Attachment 2           ,
to U-603276 -
Page 238 of 252 failure of any other safety related system from performing its safety function. The failure of the interlock will not affect any safety-related SSC, nor will it degrade the availability of the ERAT system. . The' compensatory actions have no affect on radiation levels or exposures, nor To they
  ; impact fission product barriers. These actions will not degrade, change or prevent any system from mitigating the effect of an accident or transient of any type. The compensatory actions will not affect the reliability of any SSC or the availability, performance, and reliability of the           -
RAT / ERAT systems and safey related IE 4160 bus. Therefore, this activity does not increase the probability or consequences of an' accident or malfunction of equipment imponant to safety
   . previously evaluated in the USA.R. The possibility of a complete failure of off-site power caused         a
: by operators closing both I APET4 and 1 APET14 is a credible event. The deficient interlock
   . reduces the baniers between disconnect switch 1 APET4 and 1 APET14, but compensatory actions are being implemented to replace them with others. Loss of Off-site power is a credible failure mode that has already been analyzed. There are no new failure mechanisms being initiated due to the compensatory actions. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. Technical Specifications 3.8.1.1 and 3.8.2.1 and their associated Bases -
provide operability requirements for the RAT ' and ERAT systems. In or' der for the systems to be .
operable, one off-site power source, breaker and disconnect switch to the RAT or ERAT shall be            ,
  ; feeding power to all 4.16kV buses. The availability of either off-site power source will not be affected by the compensatory actions being implemented. Therefore, this activity does not reduce a margin'of safety as defined in the basis for any Technical Specification.
NON-CONSERVATIVE ALLOWABLE VALUES FOR INSTRUMENT SETPOINTS FOR 7THE ADS INITIATION TIMERS a
   ' Activity Evaluated: Compensatory Actions for CR l-99-04-174                    Log Number: 99-094 and OE l-99-04-174-OE-1 Condition Report (CR) 1-99-04-174 documents ' discovery of a non-conservative allowable value *
   - (AV) and instrument setpoint for the Automatic Depressurization System (ADS) Initiation
   . Timers. Operability Evaluation (OE) 1-99-04-174-OE-1 was subsequently prepared, which determined that compensatory measures were necessary to support plant restart commitments.                  j The compensatory actions include revising the affected AV in the Technical Specifications Table 3.3.5.1-1 and Clinton Power Station calibration procedure 9430.30 Appendix A. This activit does not change any setpoint already established in the field and, thus, does not change any' credible failure mode.' The instrument setpoint functions in the scope of the subject activity are designed to actuate mitigating equipment should an accident or transient occur. The revised AV.is based on a 95% error probability limit at a 95% confidence value (as accepted by Regulatory Guide 1.105, Revision 2).;This activity provides additional assurance that the equipment operability and system availability assumptions utilized in plant accident analyses will.
not be impacted.~ This activity does'not degrade' design basis performance by affecting any environmental, seismic or separation design bases or criteria. Therefore, this activity does not increase the probability of an accident or malfunction of equipment important to safety previously
   . evaluated in the Updated Safety Analysis Repon (USAR). This activity does not affect overall                3

4                                                                                     Attachment 2 to U-603276 Page 239 of 252 system performance, change system response characteristics, cause system operation outside of design limits, cause system operational transients, or cause adverse system interactions. No plant equipment was replaced, removed or added as a result of this activity. Therefore, this activity does not create the possibility of an accident or malfunction of eguipment important to safety of a different type than previously evaluated in the US AR. The margin of safety, as dermed in the context of this evaluation, is considered to be the difference between the analytical limits and the design failure point for the related equipment, system or function. Since the subject activity
      ' provides an additional conservatism between the analytical limit and the Technical Specification allowable value, the margin of safety is not impacted. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
ESTABLISHMENT OF CONSERVATIVE LIMITS FOR RCIC AND HPCS SUPPRESSION POOL LEVEL HIGH TECHNICAL SPECIFICATION ALLOWABLE VALUES Activity Evaluated: OD/OE l-98-09-461                                           Log Number: 99-095 This activity establishes an administrative limit to lower the Allowable Value (AV) for Suppression Pool High Level per Operability Determination (OD) 1-98-09-461 from s 12" (High Pressure Core Spray (HPCS) Suppression Pool Level - High) and s -3" (Reactor Core Isolation Cooling (RCIC) Suppression Pool Level - High) to s 8.5" and s -6.5", respectively. The more conservative limit ensures the Suppression Pool level will not exceed the high level limits for design basis accidents and events. The Suppression Pool functions to mitigate transients and accidents that can affect the fission product barriers of the fuel, reactor and reactor coolant system, and the containment. The new limits provide ample tolerance to the Nominal Trip setpoints such that normal operation and surveillance activities will not be impacted. This activity does not change any setpoint already established in the field and, thus, does not change any           j credible failure mode. This activity ensures that the instrumentation will function to maintain        i Suppression Pool level during applicable Design Basis Accidents (DB As) at conservative values        i based on structural loading consideration. Ensuring that RCIC and HPCS suction transfer to the         ;
Suppression Pool occurs conservatively provides protection of the Reactor Pressure Vessel             i pressure control functions of the Safety Relief Valve (SRV) and related piping and spargers. The i
actual control setpoint, the operation of the HPCS and RCIC to inject during DBAs, and the integrity of the fuel remain unchanged by this activity. This activity provides additional assurance that the equipment operability and system availability assumptions utilized in plant accident analyses will not be impacted. This activity does not degrade design basis performance by
;        affecting any environmental, seismic or separatior design bases or criteria. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). This activity does not affect overall system performance, change system response characteristics, cause system operation outside of design limits, cause system operational transients, or cause adverse system interactions. No plant equipment was replaced, removed or added as a result of this activity. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the US AR. The Technical Specification Bases for RCIC Suppression Pool High level specifies that transfer occurs

Attachment 2            ,
to U.603276 Page 240 of 252 i before water level reaches the poim at which suppression pool design loads would be exceeded.
            , The Bases for HPCS Suppression Pool High level specifies that excessively high suppression pool E water level could result in the loads on the suppression pool exceeding design values should there be's blowdown of the reactor vessel pressure through the SRVsc The more conservative
            . surveillance limits for the setpoint ensures that the setpoint will be adjusted conservatively enough to affect transfer of HPCS and RCIC suction at levels that do not impact structural loadsc Since           -
           ' the change is more restrictive with respect to level, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
EVALUATION OF CONTROL ROOM HVAC SYSTEM (VC) WITHOUT VERIFICATION OF SMOKE DETECTORS Activity Evaluated hCPS Procedure 9070.04, PDR 99-0712.                          Log Number: 99-098 CPS Procedure 9070.04D001, PDR 99-0713 This activity covers the plant startup without verification of the smoke detectors' capability to automatically initiate the smoke mode of Main Control Room Heating, Ventilation, and
           . Air-Conditioning (HVAC) System (VC). The detectors will be declared nonfunctional, and the compensatory actions of the Updated Safety Analysis Report (USAR) implemented. Chapters 6 and 15 of the USAR do not address the control room HVAC smoke mode of operationc The VC system is an accident mitigation system and is not considered an initiator for any accident
           - previously evaluated in the USAR. The actions prescribed by the USAR (verify manual initiation of smoke mode for nonfunctional smoke detectors) will be verified prior to startup, thus overall system performance will not be affected. The control room HVAC smoke mode of operation has
           . no impact upon radiological considerations for the design basis accidents and transients, because a radiological accident is not assumed to occur simultaneously with a fire. Moreover, the system does not provide a direct mitigatory function for accident consequences; and this change does not            {
degrade the safety related functions of the VC system to protect the control room' operators and             I provide cooling to the control roomL This activity does not change, degrade, or prevent actions described in the accident analysis for mitigating the effect of an accident or transient. Because the safety functions of the VC system are tested via Clinton Power Station procedures 9070.02 and 9070.03, testing of the VC smoke mode after plant startup does not impair the VC safety function availability or operability, and does not degrade VC safety function performance. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the USAR. A failure of the control room                !
            -ventilation automatic smoke mode will not impair the safety related functions of the VC system.
           , Whether the VC system is operating in normal mode or in the smoke mode of operation has no c impact upon temperature control.1 A radiation signal overrides operation in normal mode or smoke mode, thus a failure of the control room ventilation smoke mode to automatically initiate
            .will not impair the function of protecting the operator from radiation during an accident.
           . Therefore, this change does not create the possibility of an accident or malfunction of equipment important to' safety of a different type than previously evaluated. The control room ventilation smoke mode of operation is not required to be operable to support any safety related or accident mitigating structure, systems, or components. The control room operators are still able to

C
;q                                                                                             Attachment 2 to U-603276 Page 241 of 252
            ' perform their necessary functions to assure the Technical Specifications Limiting Conditions for Operation, limiting safety system settings, and safety limits are complied with. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
           - INSTALLATION OF A TEMPORARY MANUAL PURGE VALVE IN OUTLET OF
: SOLENOID VALVE OFSVVC614B                                                                                   . i Activity Evaluated: - Temporary Modification 99-038 -
                                                                                              - Log Number: 99-099 The solenoid purge valve for the dehydrator of Control Room Heating, Ventilation, and Air-Conditioning (HVAC)(VC) Train B chiller (0VCl3CB) is sticking open/failing to close. _This results in a leak pa_th for refrigerant from the chiller. In order to stop the leak, and still be able to purge the non-condensable gas from the VC B chiller, Temporary Modification 99-038 installs a temporary valve on the outlet of the existing solenoid valve. The temporary valve on the outlet of the solenoid valve allows the operators to bleed off any non-condensable gas from the chiller whenever it is required. Once a new solenoid valve is procured, the valve will be replaced, and -
the Temporary Modification will be removed. The control room HVAC system, is designed to provide habitability in the control room under normal and abnonnal station conditions.
: Installation of this temporary modification will not alter the ability of the Division II VC chiller to perform its safety function nor will it change, degrade, or prevent actions that would mitigate the effects of an accident or transient. In addition, the installation of the temporary bleed offvalve will ensure that the VC chiller is run per the original design. The bleed off valve and the change to the "C"_ area operator log will ensure that the build up of non-condensable gas does not reach a       ,
level that could affect chiller performance. The installation of this temporary modification has no impact on the seismic qualification of the chiller. Therefore, this activity does not increase the probability or consequences of an_ accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR). Failure of the valve would mean a loss of freon from the machine which would, ifleft unchecked, result in a loss of the Division 11 VC chiller. The control room HVAC system is provided with redundant equipment
            ' trains te meet the single failure criteria. Loss of one Division of the VC system has already been evaluated in the ~USAR. Therefore, this activity does not create the possibility ofan accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. Technical Specification Section 3.7.4 and its associated Bases discusses the requirements to have two control room HVAC subsystems for Modes 1,2, and 3. This temporary modification does not alter those requirements. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
MAIN GENERATOR EXCITER STABILITY CALIBRATION .
Activity Evaluated: CPS' Procedure 2821.00                                        Log Number: 99-100 This activity performs a stability verification of the Alterex Exciter automatic regulator board.
This is a vendor recommended activity typically performed at other nuclear and non-nuclear sites.
In the past, this activity has not been performed at Clinton Power Station (CPS). Thus, it will be

1 Attachment 2            ,
to U-603276 Page 242 of 252 treated as a first time evolution and will be performed using CPS Procedure 2821.00, The
       ; TurbineLG:nerator is not required to support safe shutdown of the reactor or to mitigate the consequences of an accidentJ Updated Safety Analysis Report (USAR) Section 15.2.2 does discuss a Generator load reject sitization with a subsequent Turbine trip. This activity, however, qwill be performed with no potential for load rejection since the Generator will be unloaded while ithe test is being performed. This activity verifies the function of an existing design feature of the        -
Main Generator. The verification of the proper function increases the reliability of the circuit to respond to grid transients, and therefore, minimize turbine trips or transients to the plant. The activity being performed is within the design capabilities _of the Main Generator and is recommended by the vendor. This activity is performed prior to the generator being synchronized
       . to the distribution grid, when reactor power is at approximately 15% power, The power level
                                           ~
that the test will be performed at (approximately 15% power) is well within the capabilities of the turbine; bypass valves (approximately 35% power). 'All Turbine operating and protective features will be in service and furrtioning and will not be adversely affected by this activity. This activity remains bounded by the Generator Failure accidents and transients previously evaluated in
       - Chapter 15 of the USAR. Thecefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. . No equipment design 'is being changed. - All supporting systems and components will be operated within their normal operating design capabilities. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The Main Generator and Alterex are not associated with any Technical Specifications 0 This activity does not adversely impact any system or configuration that may -
        ' directly or ~ indirectly atTect any Technical Specification including all electrical power distribution
       , requirements.; Therefore, this activity does not reduce a margin of safety as defined in the basis foi any Technical SpecificationL CHANGE TO ITS BASES SR 3.3,1.1.6 - SRM/IRM OVERLAP REQUIREMENT Activity Evaluated: Technical Specification Bases Change BE-99-016                Log Number: 99-103 Technical Specification Bases Change BE-99-016 allows establishing a Source Range Monitor
       - (SRM)/ Intermediate Range. Monitor (IRM) overlap prior to reaching the upper SRM limit. The change ensures the neutron flux monitoring equipment is operated as intended by design.
Establishment of overlap is not an event initiator for any of the design basis accidents as defined in
        . Chapters 6 and 15 of the Updated Safety Analysis Report (USAR). This activity does not alter overall performance and response'of SRM and IRM instrumentation. This Bases change will not cause the SRM/IRM system to operate outside ofits design limits, cause operational transients or adverse system interaction. SRMs are not assumed 19 function during any design basis accidents or transients.' This change does not involve a hardware, software, or a setpoint change, and does not impact the neutron flux monitoring system protective functions. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The activity does not introduce any new
       . equipment failure modes for the SRM/lRM instrumentation systems. The design function of the
        - SRM/IRM overlap will be maintained by this activity. ' Also,'this activity does not make any
{ ',
4

i~                                                                                          Attachment 2 to U-603276 Page 243 of 252 physical changes to the plant or add any new equipment or components to the systems analyzed in the USAR. ' Therefore, this activity does not create the possibility of an accident or malfunction of
             . equipment important to safety of a different type than previously evaluated. No margin of safety is dermed in the Technical Specification Bases for establishing an overlap between SRM and IRM instrument channels.: The minimum required overlap is being changed by this activity; however,-
             . this is an administrative and operational margin rather than a safety margin. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REDUCE RCIC MAXIMUM OPERATING PRESSURE BAND
             '' Activity Evaluated: Operability Evaluation for CR l-99-05-124                     Log Number:' 99-104
    '    .    .The Operability Evaluation for Condition Report (CR) 1-99-05-124 re-defined the range of the operating pressure for the Reactor Core Isolation Cooling (RCIC) System. This change reduced           ,
the maximum operating pressure of the RCIC system from 1215 psia to 1118 psia in Technical
: Specification Bases B3.5.3. The RCIC system functions as a mitigatory system in response to a Design Basis Accident (DBA). This change does not alter the probability ofinadvertent RCIC
             , System operation; thus, the probability of an Inadvertent HPCS/RCIC Pump Startup (Updated Safety Analysis Report (USAR) Section 15.5.1) event is unaffected. This change can not act as an accident initiator in any other DBA defined in Chapters 6 and 15. This activity does not create
                         ~
any new system, structure or component (SSC) interfaces'that could create any harmful SSC interactions. The maximum injection pressure is a design parameter which will not impact the way the RCIC system res;ionds to any event previously evaluated. This activity does not alter any of the original design function, material, or construction standards applied to the RCIC system.
Therefore,' this activity does not increase the probability or consequences of an accident or' malfunction of equipment important to safety previously evaluated in the USAR. Changing the
               . backpressure against which the system is required to pump cannot affect an inadvertent start, induce any other credible failure mechanisms, or make any failure mechanism more likely.
Therefore, this activity does not create the possibility of an accident or malfunction of equipment '
                                 ~
               . important to safety of a different type than previously evaluated in the USAR. The functional requirement for RCIC is to be able to pump at least 600 gallons per minute (gpm) into the vessel so that challeng;s'to Emergency Core Cooling Systems are minimized. This activity does not affect that' capability; RCIC will still pump' at least 600_ gpm against the maximum reactor pressure (conservatively 1118 psia) that would exist for the most severe postulated transient for which RCIC is required to respond. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
: BYPASS THE 'A' TURBINE DRIVEN REACTOR FEED PUMP DISCHARGE VALVE
               ' 90-100% OPEN PERMISSIVE
  }
                  ~ Activity Evaluated: . Temporary Modification .99-044                           Log Number: 99-106 Temporary Modification 99-044 installs a jumper around the valve position contact for IFWOO2A
: in the "A" Turbine Driven Reactor Feed Pump (TDRFP) feedwater level controller logic. This
                                                                                                                          )

i Attachment 2            ,
to U-603276 Page 244 of 252 -
r jumper is necessary to allow providing feedwater to the reactor vessel using the "A" TDRFP and L its associated feedwater bypass valve. During startup, the position of IFW002A is closed, therefore this activity is required in order to' allow feedwater control via the "A" TDRFP flowpath. The Motor Driven Reactor Feed Pump (MDRFP)is normally used to supply feedwater
                                                         ~
during' reactor startup; however, a second flow path is desired during trouble shooting and for redundancy should the MDRFP flow path become unavailable.
: Updated Safety l Analysis Report (USAR) Chapter 15 and ~ Appendix A of Chapter 15 examines the variou's failure mode considerations for this system relative to plant safety and operational effects.
The transients or events pertinent to this activity are loss of feedwater flow and feedwater -
         - controller failure - maximum demand. The purpose of the interlock being defeated is to ensure a flow path for the "A" TDRFP when it is running and supplying feedwater flow, Strict procedural control per Clinton Power Station Procedure 3103.01, "Feedwater," will ensure this valve is
         - positioned to provide an adequate flow path when the "A" TDRFP is running. The credible failure mode for this activity is the failure of thejumper to maintain circuit continuity (open circuit). The failure to maintain circuit continuity exists in the present circui_try. The present circuitry utilizes a contact from the IFWOO2A motor operated valve rotary switch to maintain continuity in the circuit. This activity installs a jumper to maintain continuity, which has less
         ' probability of failing than an active switch due to the inherent potential for high contact resistance.
V         This jumper is installed in a part of the feedwater control circuit that would not cause a maximum -
feedwater demand signal, therefore, this activity does not have the ability to result in system configurations that would ' alter the probability of a maximum feedweter demand signal failure.
1The consequences ofloss of feedwater flow or a feedwater controller failure do not result in any pressure or temperature transients that would challenge the integrity of the fuel barrier,
_        Containment, or the pressure vessel that could increase off site radiation dose. This activity does not alter the assumed sequence, method of propagation,'or assumed failure modes for the above accidents, therefore, accident consequences that may result from this activity would be less' severe and bounded by the analysis described in the USAR. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety
          ' previously evaluated in the USAR. The materials and installation requirements for this activity are con'sistent with the qualification of the existing panel and circuits. This activity does not change the function or objective of the feedwater control system. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. This' ectivity does not alter the feedwater system configuration 'or operation in a manner that degrades system capability. This modification does
         , not invalidate any acceptance values, safety limits, limiting safety system settings, or limiting conditions for operation associated with any Technical Specifications. Therefore, this activity
         ' does not reduce a margin of safety as defined in the basis for any Technical Specification.
1 5

re                                                                                                                ]
   .,                                                                                       Attachment 2 to U-603276 Page 245 of 252
         ~ REPLACEMENT OF FUSES IN THE REMOTE SHUTDOWN PANEL FOR RHR A PUMP
         = BREAKER DC CONTROL POWER-J Activity Evaluated: ECN 31677                                                    Log Number: 99-108
!          Engineering Change Notice (ECN) 31677 replaces four MIN-10 and MIN-30 fuses in the Remote            - >
1 Shutdown Panel (RSP) with MIN-15 fuses. These fuses provide over-current protection for the Direct Current (DC) control power for the Residual Heat Removal (RHR) "A" pump 4kV
         - breaker closing and tripping circuits after switch C61-S12 is used to transfer control of the pump from the Main Control Room (MCR) to the RSP. Since these fuses are located at the RSP, they normally are not carrying any current. These fuses are only placed into service in response to an event requiring usage of the RHR "A" pump as part of safe shutdown of the plant from outside of the MCR. The drawing revision required by this design change is limited to showing the new amperage ratings for the fuses on figure E02-lRS99-101 which is incorporated by reference in
: Updated Safety Analysis Repcrt (USAR) Table 1,7-1. There is no change to the safety
         - classification of the circuit or its source at the Division I DC bus. Calculations 19-D-34 has           -
previously demonstrated the acceptability of the MIN-15 fuse ratings.
The activity only affects specific circuits associated with the RSP and the RSP is not an assumed
          ' initiator for any Design Basis ' Accident The new fuses function as well or better than the MIN-10    ,
and MIN-30 fuses with respect to current carrying ability, are rated to operate in 125 VDC circuits and have intenupting capability greater than the available fault current present at the RSP DC bus. Accordingly, there will be no change to the operation of any equipment as a result of            1 this activity. l Since this change has no negative impact on the operation of the RHR pump, it cannot impact the use of the pump to provide cooling of the vessel or the suppression pool in a post accident or safe shutdown scenario. In order for the fuse to operate and remove control
          . power from the pump breaker, there must first be a fault in the control circuit. The fuse can only respond to the fault, it cannot initiate the fault. The replacement fuses are the same physical size and are mounted in the same fuseholders as the original fuses, thus, there is no impact to the panel's seismic qualification. The internal layout and wiring of the panel is unchanged, so there is no impact to separation criteria. The fuses are passive current carrying devices that only operate
                                    ~
: in respo'nse to a fault in the connected equipment. . They can neither increase nor decrease the
         . potential for the connected equipment to experience a fault. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. Fuses can only fail in two ways; by failing to operate       1 or by operating spuriously. These failure mechanisms were present with the originally installed         j fuses and the replacements are neither more nor less susceptible to them. Therefore, this activity      I does not create the possibility of an accident or malfunction of equipment important to safety of a different tyre than previously evaluated in the USAR. This activity does not change any Technical Specification, safety limit, limiting safety system setting, or limiting condition for
;           operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for
          . any Technical Specification.
i
                                                                                                                   ]
I

Attachment 2               i to U-603276 Page 246 of 252 l
  - MAINTAIN CONDENSATE FLOW THROUGH BOTH SJAE INTERCONDENSERS AND                                            I OG RECOMBINERS Activity Evaluated! ECN 31683; USAR Change 8-319                               Log Number: 99-109 Engineering Change Notice (ECN) 31683 revises Updated Safety Analysis Repon (USAR)                        ,
Figure 10.4-5 to show condensate (CD) inlet valve ICD 006B to the 'B' Steam Jet Air Ejector               )
(SJAE) and Off-Gas (OG) recombiner as normally open. The only design basis accidents that-                I could be affected by this change are Loss of Condenser Vacuum and Loss of Feedwater Flow.
Keeping flow through both trains allows the standby recombiner to be kept preheated in a standby condition.c If a problem occurs with the on-line SJAE, the standby SJAE could be placed in
: service and possibly prevent a Loss of Condenser Vacuum event. Operational experience and initial plant testing sho'w that normal system operation is with condensate flow through both trains, and with the ability to swap SJAEs, this change actually reduces the probability of a Loss 1
  - of Condenser Vacuum accident. . Loss of Feedwater Flow could occur due to a malfunction that              l would restrict flow to the reactor or due to a pressure boundary los's. . Since this change provides for two parallel paths of condensate flow through the SJAE intercondensers and recombiners, a single failure of a valve in either train could not cause a loss of flow. The SJAE is not relied upon to mitigate the effects of an accident and has no safety f mction. While not relied upon, the             1 condensate system may be used if available to provide a water supply to the reactor to assist in mitigating accidents. This change has no adverse impact on the ability of the condensate system J
to perform this function. This change has no impact on the gaseous radwaste system integrity or
  ' function, or any fission product barriers. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously                 )
evaluated in the USAR. This change does not create the possibility of an accident or malfunction of eqwpment important to safety of a different type than previously evaluated. The systems impacted by this change are not considered imponant to safety and are not relied upon to maintain any margin of safety either in normal plant operation or under transients or accidents evaluated in the USAR. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specifications.
LOW FLOW TO RCIC LUBE OIL COOLER Activity Evaluated: Operability Evaluation for CR l-99-05-191                  Log Number: 99-111 Updated Safety Analysis Report (USAR) Section 5.4.6.2.2.2 indicates that the required cooling water flow to the Reactor Core Isolation Cooling (RCIC) lube oil cooler is 16 to 25 gallons per           j minute (gpm), also this section estimates a 0.343" orifice to achieve this flow. According to
: USAR Figure 5.4-9 sheet 2, the actual installed orifice is 0.238" which makes the actual cooling water flow rate through the RCIC lube oil cooler approximately 10 gallons per minute.
Operability Evaluation 1-99-05-191-OE-1 determined that 10 gpm nominal flow rate is adequate to maintain the turbine bearing lube oil supply temperature within limits. The RCIC system               ,
functions.as a mitigatory system in response to a Design Basis Accident (DBA). The. . mge in the RCIC lube oil cooler cooling water flow is not an accident initiator in any of the Dlw

ie                                                                                       Attachment 2 to U-603276 Page 247 of 2$2 defined in Chapters 6 and 15. The changes do not affect piping structural integrity since piping pressure and flow are not adversely affected. The only credible potential failure mode associated with the reduction in cooling water flow would be a failure of the RCIC turbine bearings with a subsequent functional loss of the RCIC system. The design function for the RCIC lube oil cooler
        . is the removal of heat from the turbine lube oiland transfer the heat to the cooling water. The heat in the lube oil comes from the RCIC turbine bearings ;vhich are cooled by the lube oil. The General Electric design specification does not invoke a specific limit for normal operation, but the bearing inlet high temperature alarm is set at 160 F. USAR Section 5.4.6.1.1.1 specifies a water temperature range of 40 to 140 F for the RCIC pump; Analysis has been' performed which shows a small increase (approximately 1 F) in the bearing lube oil return temperature (to -
approximately 151*F) based on 140 F cooling water supply at 10 gpm. This is well below the operational lube oil return limit'of 185*F and the 160 F alann setpoint. ~As such, there would be no adverse affect on the bearings as a result of the reduced cooling water flow. Since the operational limits of the bearings are met, the RCIC turbine will perform as required and there is adequate basis for reducing the lube oil cooling water flow. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the.USAR. The RCIC system continues to function as intended.
This activity does not install any new structure, system, or component or install any new system interactions. Therefore, this activity does not create the possibility of an accident or maifunction of equipment important to safety of a different type than previously evaluated in the USAR.
Technical Specifiestion Surveillance Requirements include verificatioc of system' fill, correct valve position, low pressure and high pressure operability tests and automatic actuation capability. The accident mitigation functions and performance.of the RCIC system under postulated accidents are not adversely affected. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
       . RELOCATE FRG RESPONSIBILITIES FROM ORM TO USAR; DELETE REQUIREMENT TO REVIEW TECH SPEC (TS) VIOLATIONS FOR AN UNREVIEWED SAFETY QUESTION Activity Evaluated: USAR Charg 3-326 ' .                                       Log Number: 99-117
                                 - and ORM Change 26-5 Condition Report 1-98-02-177 identified that the Operational Requirements Manual,(ORM) lists a revit.w responsibility for the Facility Review Group (FRG) that is meaningless, and therefore, has not been performed as required.: This activity revises the ORM, Updated Safety Analysis Report (USAR), and Clinton Power Station (CPS) Procedure 1001.03, " Facility Review Group,"
to accurately reflect the FRG.responsibilit ies. In a/dition, a comparison of the ORM
       - requirements was made to the USAR requirements. As a result of this review, ORM requirements over and above those described in the USAR are being transferred to the USAR to ensure that the USAR will specificallyidentify the FRG responsibilities. Also, USAR 13.4.1.d
       . (vii)is being revised to be consistent with the intent of ANSI N18.7-1976, Paragraph 4.3.4 with respect to FRG responsibilities. The FRG performs an advisory function to the Manager - CPS
                               ~
on matters related to nuclear safety. The changes involved with this activity do not reduce this e                                                                        _ _ _ _ _ _ _ _ _ _ _ . _ _

V
                ,                                                                                                     s Attachment 2            ,
to U-603276 Page 248 of 252 oversight' capability.jThis is an administrative change and 'dc :s not modify plant design or operation. _ No change is being made'to the operation of the facility or to the availability of any l equipment.LTherefore, this change does not increase the probability or consequences of an ,
             . accident or malfunction of equipment important to safety previously evaluated in the USAR. In addition, this activity does not add any new structure, systern, or component or create any new
..           : system interactions.- Therefore, this activity does not create the possibility of an accident or
[          -
z malfunction of eq'uipment important to safety of a different type than previously evaluated. This activity does not affect any Technical Specification, safety limit, limiting safety system setting, or Llimiting' condition for operation.7Therefore, this activity does not reduce a margin of safety as -
defined in the liasis for any Technical Specification.
             . ORGANIZATIONAL CHANGE, PROJECTS / CONTRACTS ORGANIZATION REPORTING RELATIONSHIP-e
              ' ActivityEvaluated: USAR Change 8-330                                            Log Number: 99-120-Updated Safety Analysis Report (USAR) Change 8-330 establishes the organization of Projects / Contracts including the position'of Manager - Projects / Contracts reporting to the Vice .
President / Chief Nuclear Officer. This position and the associated organization will be responsible for the management, ovcesight end coordination of modification and project implementation at Clinton Power Station (CPS). This USAR Change also includes the creation of the position of
             ' Director - Work Support and the deletion of the position of Director - Contracts' The Director -
              . Work Suppoit will report to the Manager - Work Management. The re-assignment of the
             ~ functions and responsibilities of the Director - Contracts to the new position, Manager -
: m.          Projects / Contracts, and the creation of Director - Work oopport, does not result in a reduction of commitments or effectiveness of the' CPS organization and its responsibilities as previously accepted by the _ Nuclear Regulatory Commission.' This is' an administrative organizational change and does not modify plant design or operation of any structure,' system, or component. No change is being made to the operation of the facility or to the availability of any equipment.
Therefore, this change does not increase the probability or consequences of an accident or inalfanction of eauipment important to safety previously evaluated in the USAR. In addition, this
: activity does not add any new structure, system, or component or create any new system interactions. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The management
_ responsibilities / requirements delineated in Section 5.0 of the Technical Specifications are
             ! associated with the Plant Manager and Operations personnel. The programs and reporting requirements addressed in Section 5.0 are not affected by the creation of the new organization, t          .the creation of the new positions, or the deletion of the Director - Contracts position._ Therefore, this activity does 'not reduce a margin of safety as defined in the basis for any Technical Specification.;
1

     ~
l 0'                                                                                  Attachment 2 to U-603276 Page 249 of 252 ADDITION OF SILT AND BIO DISPERSANT CHEMICALS TO THE CW SYSTEM l
Activity Evc'uated: USAR Change 8-333                                          Log Number: 99-122 At the time of this evaluation, the Updated Safety Analysis Report (USAR) limited the chemicals added to the Circulating Water (CW) system to just those chemicals needed for corrosion control and scale inhibition. This activity changes USAR Section 10.4.5.2 to allow the use of analyzed I      chemical addition regimes for the control of biological growth and suspended solids. Also, this activity adds Depositrol PY5200 and Spectrus BD1500 to the CW system to allow lowering of Main Condenser differential pressure. This activity has the potential to impact Main Turbine Trip, Loss of Main Condenser Vacuum, Operation with Leaking Condenser Tubes, the Fire Hazards Analysis, and Mai6 Control Room Habitability. Clinton Power Station procedures limit the use of chemicals in the CW system to those evaluated and found acceptable and impose control over the water treatment vendor. The presence of the chemicals in any of the systems does not increase material degradation, because the chemicals are compatible with the materials in the condenser, condensate and feed system, and the reactor. The chemical addition rate is controlled to prevent chemical shock and the water treatment chemicals are neither volatile or hazardous. The use of        )
the new chemicals does not result in system configurations or tube failures that challenge the operation of any safety system, or impair system reliability when relied upon for system protection. The principal equipment affected by the new chemicals consist of CW piping, condenser water boxes, and condenser tubes. The Main Condenser and CW System are not required to affect or support safe shutdown of the reactor or to support the operation of reactor     ,
safety features. Thus, the affects of the chemicals upon these components have no direct affect upon equipment impcrtant to safety. This activity will not change, degrade, or prevent actions described or assumed in the accident analysis for mitigating the effects of any transient. Also, the  !
activity cannot result in a change in the radiological consequences of any accidents or transients, the radiological release limits associated with any accidents or transients will not be exceeded, and no fission product barriers will be challenged. Therefore, this activity does not increase the        l probability or consequences of an accident or malfunction of equipment important to safety            l previously evaluated in the USAR. The type of accidents and malfunctions of equipment important to safety which could result from this activity have already been addressed in the USAR. The plugging of the Main Condenser from particulates, assuming a worst case scenario would most likely not cause immediate loss of condenser vacuum or a trip of the main turbine.
Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. This activity does not affect any Technical Specification, safety limit, limiting safety system setting, or limiting condition for operation. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
l l
L

Attachment 2                j l
to U-603276 Page 250 of 252
: l. -
(
CLARIFY OPERATION OF THE RHR HEAT EXCHANGER VALVES Activity Evaluated: USAR Change 8-337                                            Log Number: 99-125             i I
l This activity incorporates clarifications / corrections to the Updated Safety Analysis Report
:      ' (USAR) for the containment spray logic to the valves around the Residual Heat Removal (RHR) l        heat exchanger. This clarifies that if the RHR heat exchanger inlet or outlet valves are not fully
: l.       open, the bypass' valve will not automatically close'. USAR Change 8-337 revises this logic on drawing E02-RH99-019, revision K, which is listed in Table 1.7-1 of the USAR. The RHR heat exchanger is credited with mitigating many of the accident scenarios listed in Chapters 6 and 15.
l        The heat exchanger is required for containment and shutdown cooling. As such, the valves surrounding the RHR heat exchanger are~also credited, either actively or passively, with               -
mitigating numerous' accident scenarios. Howe'ver, the RHR heat exchanger and its associated l
valves are only credited with initiating the following events: inadvertent operation of shutdown
!      : cooling and flooding scenarios involving the service water and RHR lines. The logic associated L        with the automatic losing of the RHR side bypass valve is only active on initiation of containment
!        spray. . If a containment spray signal is present, the operation of the RHR heat exchanger is desirable and, therefore, does not constitute inadvertent operation. The clarifications / corrections
: of the valve logic implemented by this activity does not provide a larger flow path or prevent the isolation of a flow path. This activity does not affect the pressure boundary components in any way, does not affect the availability / effectiveness of overpressure protection, and does not subject the system to transients or conditions not previously considered in the system design. This activity does not involve any changes in materials, any new modes of system operation, or any change in seismic or environmental qualification of equipment. Therefore, this activity does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This activity does not introduce any new operating .
modes, failure mechanisms, or single failure points. Nor does this activity introduce any new
                              ~
structure, system or component. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR. The bases for the Technical Specifications regarding various modes of RHR tiescribe that the RHR heat exchangers are required for operability of the containment spray mode and the suppression pool cooling mode. The clarification of the interlocks does not conflict with any discussion provided in the bases to these Technical Specifications. The interlock requirement to have the RHR heat exchanger inlet and outlet valves full open prior to closing the bypass 2ve helps to ensure that the required flow established by Technical Specification Surveillance Requirements 3.5.1.4 and 3.6.1.7.2 are met. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical. Specification.
i

          - w.
i:                                                                                         Attachment 2 to U-603276 Page 251 of 252
               . ORGANIZATIONAL CHANGE, DIRECTOR-CORRECTIVE ACTION REPORTING RELATIONSHIP Activity EvaluatedF USAR Change 8-339                                           Log Number: 99-128 Updated Safety Analysis Report (USAR) Change 8-339 changes the reporting relationship of the          -
               ' Director - Corrective Action from the Manager - Nuclear Safety and Performance Improvement to the Manager - Clinton Power Station. The specific duties and responsibilities of the Director -
Corrective Action and the Corrective Action Program are unchanged. This is an administrative
                ' organizational change and does not modify plant design or operation of any structure, system, or component. No change is being made to the operation of the facility or to the availability of any equipment. Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the USAll In addition, this activity does not add any new structure, system, or componentLand there are no new failure modes or event sequences introduced. Therefore, this activity does not create the
                . possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated. The management responsibilities / requirements delineated in Section 5.0 of the Technical Specifications are associated with the Plant Manager and Operations personnel.
This change of reporting relationship does not affect those responsibilities. Therefore, this activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
REVIS'E CHILLER MONITORING INFORMATION
               ' Activity Evaluated: USAR Change 8-341                                            Log Number: 99-129 The Updated Safety Analysis Repon (USAR) indicates that the differential pressure of the water passing through the condenser and evaporator of a variety of chillers are periodically monitored to ensure the chiller is functioning properly. This statement is not correct! there is no instrumentation to verify these differential pressures. The verification of these parameters would
               ' be beneficial, but it is not required to verify the chillers are functioning properly. Since there are (no existing means to take these measurements and the installation of these means would require            ,
plant modifications of very small benefit versus cost, USAR Change 8-341 removes the ponion of         !
Lt he statements concerning the monitoring of water differential pressures. The other chiller parameters discussed, which are monitored periodically, will remain. None of the subject systems,      j
                -including their chillers, are the initiator of any accidents evaluated in Chapters 6 or 15 of the USARc This activity does not make any physical changes to any Stmeture, System, or Component (SSC); The differential pressure of the water across the evaporator and the condenser are not the critical parameters for determining proper chiller operation. The chiller parameters'which are periodically monitored are sufficient to insure proper operation of the chillers and their supporting roles for other systems. This activity does not reduce the monitoring actually being performed to verify proper operation of the chillers, it simply removes the incorrect   i information of the parameters which are not currently moaltored due to equipment design. This          i activity does not impact the monitoring capabilities that provide early detection of malfunction of
               ~t he chillers and does not change the operation of any plant SSC. The Control Room Heating, k ;-
e

w Attachment 2            ,
to U-603276 Page 252 of 252
                                                                  ~
      - Ventilation and. Air-Conditioning (HVAC)'(VC) system does not directly mitigate the effects of.
any accident, relative to the radiological dose received by the general public; however, it does mitigate the radiological dose received by the Operations personnel in the Main Control Room I
       ~ during an accident. The parameters being removed from the USAR are not required to be monitored to ensure proper operation of the VC chiller. ' Additionally, the VC chiller is not
      ' needed to' meet the radiological dose mitigation function. Therefore, this activity does not                -
: increase the probability or consequences of an accident or malfunction of equipment important to
      ' safety previously evaluated in tb USAR. The activity does not change the operation of or -
physical condition of any plant SSC. The VC chiller and the non-safety related chillers will' continue to be ; operated and tested as before. The monitoring / testing is more than sufficient to verify the operability of the chillers. Therefore, this activity does not create the possibility of an accident or malfunction of equipment imponant to safety of a different type than previously
       . evaluated in the USAR. Technical Specification 3.7.4 is related to the VC chiller. The VC chiller
        ~is not needed to support the VC function of controlling radiological dose. The non-safety related chillers do not have any Technical Specifications associated with them. The Surveillance -
Requirement for VC chiller operability does not rely on the monitoring of the differential pressure of water flow through the condenser or evaporator of the chiller. The bases of 3.7.4 does not mention specific parameters of the chiller to be tested or monitored in support of the Surveillance Requirement _. Therefore, this activity does not reduce a margin of safety as defined in the basis
   ^
      - for any Technical Specification.
ORGANIZATIONAL CHANGE, PLANT SUPPORT SERVICES REPORTING
       ; RELATIONSHIP-.
Activity Evaluatedi USAR Change 8-342                                            Log' Number: 99-130
      < This activitp deletes supervisor positions depicted on Updated Safety Analysis Report (USAR)
Figure 13.1-1 under the Director - Plant Support Services, as these positions are below the level
       ; of detail required by Regult. tory Guide 1.70, Revision 3. USAR Change 8-342 changes the reporting relationship of the Director - Plant Support Services from the Manager - Clinton Power.
Station to the Manager - Nuclear Support. - The specific duties and responsibilities of the
       . Director - Plant Suppon' Services are unchanged. This is an administrative organizational change and does not modify plant design or operation of any structure, system, or component. No change is being made to the operation of the facility or to the availability of any equipment.
Therefore, this change does not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. In addition, this y      , activi.ty does not add any new structure, system, or component and there are no new failure
       . modes or event sequences introduced. _ Therefore, this activity does not create the possibility of an
       .' accident or malfunction'of equipment important to safety of a different type than previously c evaluated in the USAR. The management responsibilities / requirements delineated in Section 5.0
       - _of the Technical Specifications are associated with the Plant Manager and Operations personnel.
This change of reporting relationship does not affect those responsibilities. Therefore. this
: activity does not reduce a margin of safety as defined in the basis for any Technical Specification.
S i
I

        *,I l N y     .
Attachment 3
       ,; 4                 i                                                                                        to U-603276 Page i of 29 '
 ,        -;. u ;-
        -       '                 4
   , :j .. -
s i
   /                    ,.
/           .
                      .:e .                 : ....
                              -'Aftachment 3 contains the foIlow'        m g information:
(1. : Filing information to aid in updating the Clinton Power Station Updated Safety -
EAnaly~ sis Report (USAR).. This information is not required to be filed in the USAR binders.
                                          ;  2.1 Revised pages for Revision 8 to the USAR, inc!uding the Effective Page Listing.
(.          a 4
L
                                       ..s6 O
e W
b

Attachtinctil 3     .
to U-603276
* Page 2 of 29 Clinton Power Station USAR Revision 8 Filing Instructions REMOVE / DESTROY:                          l INSERT:
i Volurite i CPS-USAR Effective Page Listing,             CPS-USAR EfTective Page Listing, Rev. 8, dated February 1999                  Rev. 8, dated August 1999 1.1-1-1.1-2                                 1.1-1 (9/90)- 1.1-2 (7/98) 1.2 1.2-20                              1.2-19 (8/99) - 1.2-20 (Rev. 0) 1.2 1.2-22                              1.2-21 (1/95)- 1.2-22 (8/99) 1.2 1.2-28                              1.2-27 (5/97)- 1.2-28 (8/99) 1.2 1.2-30                              1.2-29 (8/99)- 1.2-30 (8/99) 1.2 1.2-34                              1.2-33 (8/99)- 1.2-34 (3/97) 1.2 1.2-36                              1.2-35 (11/95)- 1.2-36 (8/99)
Fig. 1.2-3 S 1                              Fig.1.2-3 S 1 (8/99)
Fig.1.2-4                                   Fig.1.2-4 (8/99)
Fig.1.2-5                                   Fig.1.2-5 (8/99)
Fig.1.2-6                                   Fig.1.2-6 (8/99)
Fig.1.2-9                                   Fig.1.2-9 (8/99)
Fig.1.2-15                                  Fig.1.2-15 (8/99)
Fig. l.2-17                                 Fig.1.2-17 (8/98) 1.3 1.3-8                                1.3-7 (6/97)- 1.3-8 (12/97) 1.3 1.3-10                               1.3-9 (6/97)- 1.3-10 (11/97) 1.6 1.6-4                               1.6-3 (Rev. 0) - 1.6-4 (12/97) 1.6 1.6-6                               1.6-5 (12/97)- 1.6-6 (12/97)
_L6 1.6-8                                1.6-7 (12/97)- 1.6-8 (12/97)                   _
l .7 1.7-2                             1.7-1 (1/98)- 1.7-2 (8/99) 1.7-3 1.7-4                                 1.7-3 (2/98)- 1.7-4 (8/99) 1.7 1.7-6                               1.7-5 (8/99)- 1.7-6 (8/99) 1.7 1.7-8                               1.7.-7 (8/99) - 1.7-8 (8/99) 1.7 1.7-10                              1.7-9 (8/99)- 1.7-10 (8/99) 1.7 1.7-12                             1.7-11 (8/99)- 1.7-12 (8/99) 1.7 1.7-14                             1.7-13 (8/99)- 1.7-14 (8/99) 1.7 1.7-16                             1.7-15 (8/99)- 1.7-16 (8/99) 1.7 1.7-18                             1.7-17 (8/99)- 1.7-18 (8/99) 1.7 1.7-20                             1.7-I9 (8/99)- 1.7-20 (8/99) 1.7 1.7-22                             1.7-21 (8/99)- 1.7-22 (8/99) 17-235 1.7-24                               1.7-23 (8/99)- 1.7-24 (2/98) 1.7 1.7-26                              1.7-25 (2/98)- 1.7-26 (2/98) 1.7-27 1.7-28                               1.7-27 (2/98)- 1.7-28 (8/99)
(                                                                                                       l

Alt.hhnient 3 to U-603276 Page 3 of 29 1.7-29 (2/98)- 1.7-30 (2/98)
[d \    1.7-29-1.7-30 1.7-31 -1.7-32           1.7-31 (2/98)- I.7-32 (2/98) 1.7 1.7-34          1.7-33 (8/99)- 1.7-34 (8/99) 1.7-35-1.7-36            1.7-35(8/99)- 1.7-36(8/99) 1.7-37 -1.7-38           1.7-37 (8/99)- 1.7-38 (8/99) 1.7-39-1.7-40            1.7-39 (2/98)- 1.7-40 (2/98) 1.7-41 -1.7-42           1.7-41 (2/98)- 1.7-42 (2/98) 1 7-43 -1.7          1.7-43 (2/98)- 1.7-44 (8/99) 1.7-45-1.7-46            1.7-45 (8/99)- 1.7-46 (2/98) 1.7-47-1.7-48            1.7-47(2/98)- 1.7-48 (2/98) 1.7 1.7-50          1.7-49 (8/99) - I c7-50 (8/99) 1.7 1.7-52          1.7-51 (8/99)- 1.7-52 (8/99) 1.7 1.7-54      y   1.7-53 (8/99)- 1.7-54 (8/99) 1.7 1.7-56          1.7-55 (2/98)- 1.7-56 (2/98) 1.7 1.7-58          1.7-57 (2/98)- 1.7-58 (2/98) 1.7 1.7-60          1.7-59 (2/98)- 1.7-60 (2/98)
T7 1.7-62            1.7-61 (8/99)- 1.7-62 (2/98) 1.7 1.7-64          1.7-63 (8/99)- 1.7-64 (8/99) 1.7 1.7-66          1.7-65 (8/99)- 1.7-66 (2/98) 1.7 1.7-68          1.7-67 (2/98)- 1.7-68 (2/98) p       1.7-69 -1.7-70           1.7-69 (2/98)- 1.7-70 (2/98) h       1.7-71 -1.7-72 1.7 1.7-74 1.7-71 (2/98)- 1.7-72 (8/99) 1.7-73 (8/99)- 1.7-74 (2/98)                 {
l.7 1.7-76          1.7-75 (8/99)- 1.7-76 (8/99) l.7 1.7-78          1.7-77 (2/98)- 1.7-78 (2/98) 1.7 1.7-80           1.7-79 (2/98)- 1.7-80 (2/98) 1.7 1.7-82          1.7-81 (2/98)- 1.7-82 (2/98) 1.7 1.7-84          1.7-83 (2/98)- 1.7-84 (2/98) 1.7-85 -1.7-86          1.7-85 (2/98)- 1.7-86 (2/98) 1.7 1.7-88         1.7-87 (2/98)- 1.7-88 (8/99) 1.7-89 -1.7-90          1.7-89 (8/99)- 1.7-90 (8/99) 1.7 1.7-92         1.7-91 (8/99)- 1.7-92 (2/98) 1.7 1.7-94         1.7-93 (8/99)- 1.7-94 (8/99) 1.7-95 -1.7-96          1.7-95 (8/99)- 1.7-96 (8/99) 1.7 1.7-98         1.7-97 (2/98)- 1.7-98 (2/98) 1.7-99 -1.7-100         1.7-99 (2/98)- 1.7-100 (8/99) 1.7-101 - 1.7-102       1.7-101 (8/99)- 1.7-102 (8/99) 1.7-103 - 1.7-104       1.7-103 (2/98)- 1.7 104 (2/98) 1.7-105 - 1.7-106       1.7-105 (2/98)- 1.7-106 (8/99) 1.7-10'it - 1.7-108     1.7-107 (8/99)- 1.7-108 (8/99) 1.7-109 - 1.7-110       1.7-109 (2/98)- !.7-110 (2/98) 1.7-111 (2/98)- 1.7 112 (2/98)
[]
% ./
1.7-111 -1.7-112

Anachment 3   ,
to U-603276 Pat;c 4 of 29 1.7-113 - 1.7-114 1.7-1I3 (8/99)- 1.7-114 (8/99) 1.7-115 - 1.7-116 1.7-115 (8/99)- 1.7-116 (8/99) 1.7-117 - 1.7-118 1.7-117 (8/99)- 1.7-118 (2/98) 1 7-119 - 1.7-120 1.7-119 (2/98)- 1.7-120 (8/99) 1.7-121 - 1.7-122 1.7-121 (8/99)- 1.7-122 (8/99) 1.7-123 - 1.7-124 1.7-123 (8/99)- 1.7-124 (8/99) 1.7-125 - 1.7-126 1.7-125 (8/99)- 1.7-126 (8/99) 1.7-127 - 1.7-128 1.7-127(8/99)- 1.7-128 (8/99) 1.7-129 - 1.7-130 1.7-129 (2/98)- 1.7-130 (2/98) 1.7-131 - 1.7-132 1.'7-131 (2/98)- 1.7-132 (2/98) 1.7-133 - 1.7-134 1.7-133 (2/98)- 1.7-134 (8/99) 1.7-135 - 1.7-136 1.7-135 (2/98)- 1.7-136 (2/98) 1.7-l 3 7 - 1.7-138 1.7-137 (2/98)- 1.7-138 (2/98) 1.7-139 1.7-140 1.7-139 (2/98)- 1.7-140 (2/98) 1.7-141 - 1.7-142 1.7-141 (8/99)- 1.7-142 (2/98) 1.7-143 - 1.~t-144 1.7-143 (2/98)- 1.7-144 (2/98) 1.7-145 - 1.7-146 1.7-145 (2/98)- 1.'7-146 (2/98) 1.7-147- 1.7-148 1.7-147(2/98)- 1.7-148 (8/99) 1.7-149- 1.7-150 1.7-149 (8/99)- 1.7-150 (8/99) 1.7-151 - 1.7-152 1.7-151 (1/98)- 1.7-152 (1/98) 1.7-153 - 1.7-154 1.7-153 (1/98)- 1.7-154 (1/98) 1.7-155 - 1.7-156 1.7-155 (1/98)- 1.7-156 (1/98) 1.7-157- 1.7-158 1.7-157 (1/98)- 1.7-158 (1/98) 1.7-159- 1.7-160 1.7-159 (1/98)- 1.7-160 (8/9'')
1.7-161 - 1.7-162 1.7-161 (8/99)- 1.7-162 (1/98) 1.7-163 - 1.7-164 1.7-163 (1/98)- 1.7-164 (1/98) 1.7-165 - 1.7-166 1.7-165 (1/98)- 1.7-166 (1/98) 1.7 167 - 1.7-168 1.7-167 (1/98)- 1.7-168 (1/98) 1.7-169 - 1.7-170 1.7-169 (1/98)- 1.7-170 (8/99) 1.7-171 - 1.7-172 1.7-171 (8/99)- 1.7-172 (8/99) 1.7-173 - 1.7-174 1.7-173 (8/99)- 1.7-174 (8/99) 1.7-175 - 1.7-176 1.7-175 (8/99)- 1.7-176 (8/99) 1.7-177 - 1.7-178 1.7-177 (8/99)- i.7-178 (8/99) 1.7-179 - 1.7-180 1.7-179 (8/99)- 1.7-180 (8/99) 1.7-181 -1.7-182 1.7-l81 (8/99) - 1.7-182 (8/99) 1.7-183 - 1.7-184 1.7-183 (8/99)- 1.~-184 (8/99) 1.7-185 - 1.7-186 1.7-185 (8/99) -1.7'l86 (8/99) 1.7-187- 1.7-188 i.7-187 (8/99)- 1.7-188 (8/99) 1.7-189 - 1.7-190 1.7-189 (8/99). ' 7-190 (1/98) 1.7-191 - 1.7-192 1.7-191 (1/98)- 1.7-192 (1/98) 1.8 1.8-2         1.8 1 (Rev. 0) - 1.8-2 (8/99) 1.8 1.8-4         1.8-3 (8/99)- 1.8-4 (8/99)

{,                                                                                 _
Attachment 3 w                                                                to U4>0327(,
Pat;c 3 of 29
,[]       1.8 1.8-6                 1.8-5 (8/99)- 1 8-6 (2/99) iV        1.8 1.8-8                 1.8-7 (8/99)- 1.8-8 (8/95) 1.8 1.8-22               1.8-21 (3/98)- 1.8-22 (8/99) 1.8 1.8-40               1.8-39 (Rev. 0) - 1.8-40 (8/99) 1.8 1.8-44              1.8-43 (8/99) - 1.8-44 (Rev. 0) 1.8 1.8-46           .
1.8-45 (Rev. 0) - 1.8-46 (2/99) 1.8 1.8-48              1.8-47 (8/99)- 1.8-48(8/99) 1.8-48a - 1.8-48b            1.8-48a (8/99) - 1.8-48b (8/99)
_l.8 1.8-56              1.8-55 (2/99) - 1.8-56 (Rev. 0) 1.8 1.8-62              1.8-61 (8/99) - 1.8-62 (Rev. 0) 1.8 1.8-68              1.8-67 (Rev. 0) - 1.8-68 (8/99) 1.8 1.8-70              1.8-69 (8/99) - 1.8-70 (Rev. 0) 1.8 1.8-90              1.8-89 (Rev. 0) - 1.8-90 (3/98) 1.8 1.8-96              1.8-95 (8/99) 96 (Rev. 0) 1.8-101 - 1.8-102            1.8-101 (Rev. 0)- 1.8-102 (8/99) 1.8-111 - 1.8-112            1.8-111 (3/97)'- 1.8-112 (8/99) 1.8-123 - 1.8-124            1.8-123 (8/99) - 1.8-124 (Rev. O) 1.8-125 - 1.8-126            1.8-125 (9/91)- 1.8-126 (8/99) 1.8-127 - 1.8-128            1.8-127 (11/97) - 1.8-128 (Rev. 0) 1.8-131 - 1.8-132            1.8-131 (8/99)- 1.8-132(8/99) 1.8-133 - 1.8-134            1.8-133 (Rev. 0)- 1.8-134 (2/99)
, C,      1.8-157 - 1.8-158            1.8-157 (2/99) - 1.8-158 (Rev. 0) 1.8-167 - 1.8-168            1.8-167 (8/99)- 1.8-168 (8/99) 1.8-171 - 1.8-172            1.8-171 (2/99)- 1.8-172 (8/99) 1.8-173 -1.8-173a            1.8-173 (11/94)-1.8-1733 (8/99) 1.8-173b - 1.8.17_3c         1.8-173b (8/99) - 1.8-173c (8/99) 1.8-173d - 1 [ '74      ,    1.8-173d (8/99) - 1.8-174 (Rev. 0) 1.8-187 - 1.8-188            1.8-187 (Rev. 5) - 1.8-188 (7/98) 1.8-201 - 1.8-202            1.8-201 (8/98) - 1.8-202 (Rev. 0)
Fig. l.91 S 1               Fig.1.9-1 S 1 (8/99) 1.10 1.10-6             1.10-5 (8/99) - 1.10-6 (Rev. 0) 2-xin xiv               2-xiii (Rev. 0) xiv (8/99) 2-xxi xxii              2-xxi (Rev. 0) xxii (8/99)
Fig. 2.1-7                  Fig. 2.1-7 (8/99) 2.2 2.2-2               2.2-1 (8/99)- 2.2-2 (8/99) 2.2-5 -2.2-6                2.2-5 (8/99)- 2.2-6 (8/99) 2.2 2.2-8        _,
2.2-7 (8/99) - 2.2-8 (9/92) 2.2-10a - 2.2-10b           2.2-10a (l/98) - 2.2-10b (2/99) l         2.2 2.2-12             2.2-1 I (3/98) - 2.2-12 (Rev. 0) 2.2 2.2-14             2.2-13 (3/98) - 2.2-14 (Rev. 0) 2.2 2.2-18             2.2-17 (9/92)- 2.2-18 (3/98)
(^]
    %j 2.2 2.2 20             2.2-19 (Rev. 0) - 2.2-20 (2/99) l I
t l
4

Attacluncut 3
                                                                              ~
to U-603276 Page 6 of 29 2.2 2.2-22              2.2-21 (S/99)- 2.2-22 (8/99) 2.2 2.2-24              2.2-23 (Rev. 0) - 2.2-24 (8/99)              /
2.2 2.2-26              2.2-25 (8/99) - 2.2-26 (Rev. 0) 2.2 Blank               2.2-27 (3/98) - Blank 2.2 Blank               2.2-28 (3/98) - Blank Fig. 2.2-1                   Fig. 2.2-1 (8/99)
Fig. 2.2-2                  Deletion Page (8/99)
Fig. 2.2-4                  Fig. 2.2-4 (8/99)
Fig. 2.2-5                  Deletion Page (8/99) 2.4-29 -2.4-30              2.4-29 (9/89)- 2.4-30 (8/99) 2.4-39 -2.4-40              2.4-39(9/91)-2.4-40(2/99) 2.4 2.4-90             2.4-89 (Rev. 0) - 2.4-90 (8/99)
: Volume 2.
2.5 2.5-78           2.5-77 (9/90)-       >-78 (7/98)
                      ; Volume 31 3-xxvii xxviii        3-xxvii (8/99) xxviii (Rev. 0) 3-il 1                1-xlix (8/99) 1 (Rev. 0) 3.1 3.1-42             ,1-41 (8/99) - 3.1-42 (Rev. 0) 3.1-49 -3.1-50              i 1-49 (2/98) - 3.1-50 (Rev. 0) 3.1 3.1-52              i 1-51 (Rev. 0) - 3.1-52 (2/99) 3.2 3.2-2              3.2-1 (9/89)- 3.2-2 (2/99) 3.2 3.2-4              3.2-3 (2/99)- 3.2-4 (2/99) 3.2 3.2-6              3.2-5 (2/99) - 3.2-6 (Rev. 0) 3.2 3 2-16            3.2-15 (5/95)- 3.2-16 (8/99) 3.2-17 -3.2.I8             3.2-17 (8/99) - 3.2-18 (Rev. 0) 3.2 3.2-24            3.2-23 (2/99)- 3.2-24 (9/94) 3.2 3.2-27a          3.2-27 (2/99) - 3.2-27a (Rev. 0) 3.2-35-3.2-36             3.2-35 (9/90)-3.2-36 (8/99) 3.2-47 -3.2-48            3.2-47 (8/99) - 3.2-48 (Rev. 5) 3.6 3.6-50           3.6-49 (3/98) - 3.6-50 (Rev. 0)
Fig. 3.6-1 S 3            Fig. 3.6-1 S 3 (3/98)
Fig. 3.6-1 S 6            Fig. 3.6-1 S 6 (2/99)
                                                                                )
Fig. 3.6-1 S 14           Fig. 3.6-1 S 14 (8/99)                              l Fig. 3.6-1 S 16           Fig. 3.6-I S 16 (12/97)
Fig. 3.6-1 S 17           Fig. 3.6-1 S 17 (2/99)
Fig. 3.6-I S 25           Fig. 3.6-1 S 25 (8/99) l Fig. 3.6-1 S 30           Fig. 3.6-1 S 30 (12/97)                             '
Fig.7.6-1 S 36            Fig. 3.6-1 S 36 (8/99)                              '

                                                                                    ---._....i" Att.alurictil 3 to U 603276 l                                                                   Page 7 cf'29
     ~
Fig. 3.6-1 S 47               Fig. 3.6-1 S 47 (8/99) l        Fig. 3.6-1 S 48              Fig. 3.6-1 S 48 (8/99)
Fig. 3.6-1 S 49              Fig. 3.6-1 S 49 (8/99)
TVolumb'4 ?                               .
Fig. 3.61 S 54               Fig. 3.6-1 S 54 (12/97)
Fig. 3.6-1 S 57              Fig. 3.6-1 S 57 (8/99) .
Fig. 3.6-1 S 58              Fig. 3.6-1 S 58 (8/99)
Fig. 3.6-1 S 60              Fig. 3.6-1 S 60 (8/99)
Fig. 3.6-1 S 61              Fig. 3.6-1 S 61 (8/99)
Fig. 3.6-1 S 62              Fig. 3.6-1 S 62 (8/99)
Fig. 3.6-1 S 63              Fig. 3.6-1 S 63 (8/99)
Fig. 3.6-1 S 64              Fig. 3.6-1 S 64 (8/99)
Fig. 3.6-1 S 68              Fig. 3.6-1 S 68 (3/98)
Fig. 3.6-1 S 69              Fig. 3.6-1 S 69 (2/99)
Fig.' 3.6-1 S 73 -           Fig. 3.6-1 S 73 (8/99)
Fig. 3.6-1 S 75              Fig. 3.6-1 S 75 (2/98)
Fig. 3.6-1 S 79              Fig. 3.6-1 S 79 (3/98)
Fig. 3.6-1 S 80              Fig. 3.6-1 S 80 (3/98)
Fig. 3.6-1 S 82              Fig. 3.6-1 S 82 (7/98)
Fig. 3.6-1 S 84              Fig. 3.6-1 S 84 (8/99)
Fig. 3.6-1 S 87              Fig. 3.6-1 S 87 (2/99)
Fig. 3.6-1 S 90A             Fig. 3.6-1 S 90A (3/98)
B3.6 B3.6-20            B3.6-19 (2/99) - B3.6-20 (2/99)
B3.6 B3.6-24            B3.6-23 (8/99) - B3.6-24 (Rev. 5)
D3.6 D3.6-6              D3.6-5 (Rev. 0) - D3.6-6 (2/99)
D3.6 D3.6-8              D3.6-7 (2/99) - D3.6-8 (2/99)                    .
D3.6 D3.6-16            D3.6-15 (4/96) - D3.6-16 (2/99)
D3.6-31                      D3.6-31 (2/99)
Fig. D3.6-88                 Fig. D3.6-88 (8/99)
Fig. D3.6-90                 Fig. D3.6-90 (8/99)
Fig. D3.6-94                 Fig. D3.6-94 (8/99) 5 "W s             '
Woldme% 0 Fig. 3.7-65                  Fig. 3.7-65 (8/99)
(        3.9 3.9-6                3.9-5 (8/99) - 3.9-6 (Rev. 0)
'        3.9 3.9             3.9-69 (8/99) - 3.9-70 (Rev. 0) 3.9-105 - 3.9-106.           3.9-105 (2/99) - 3.9-106 (Rev. 0) 3.9-119 - 3.9-120            3.9-119 (8/99) - 3.9-120 (Rev. 0) 3.9-123 - 3.9-124-           3.9 123 (8/99)- 3.9-124 (10/94) l:

Attachment 3      ,
10 U-603276 P:.rc 8 of 29 3.9-125 - 3.9-126 3 9-125 (l0/94)- 3.9-126 (8/99)~'
3.9-129 -3.9-130                                                         #
3.9-129 (8/99) - 3.9-130 (Rev. 0)~'
3.9-137 - 3.9 138 3.9-137 (Rev. 0) - 3.9-138 (8/99) 3.9-263 - 3.9-264 3.9-263 (5/97)- 3.9-264 (8/99) 3.9-265 - 3.9-266 3.9-265 (8/99)- 3.9-266 (9/89) 3.9-267 - 3.9-268 3.9-267(8/99)- 3.9-268 (9/89)
A3.9-iii- A3.9-iv A3.9-iii (Rev. 0) - A3.9-iv (8/99)
A3.9 A3.9-14 A3.9-13 (Rev. 0)- A3.9-14 (8/99)
A3.9 A3.9-16 A3.9-15 (8/99)- A3.9-16 (8/99)
A3.9 A3.9-18 A3.9-17 (8/99) - A3.9-18 (Rev. 0)
A3.9 19 - A3.9                         A3.9-19 (8/99) - A3.9-20 (Rev. 0)
A3.9 A3.9-24 A3.9-23 (Rev. 0)- A3.9-24 (8/99)
A3.9 A3.9-34 A3.9-33 (8/99)- A3.9-34 (8/99)                         l A3.9 A3.9-36 A3.9-35 (8/99) - A3.9-36 (8/99)
A3.9 A3.9-38 A3.9-37 (8/99) - A3.9-38 (8/99)
A3.9 A3.9-40 A3.9-39 (8/99) - A3.9-40 (8/99)
Fig A3.9-10            Fig. A3.9-10 (8/99)
Fig. A3.9-11           Fig. A3.9-11 (8/99)
Fig. A3.9-15           Deletion Page (8/99)
Fig. A3.9-16          Not Applicable                                            !
Fig. A3.9-17          Not Applicable Fig. A3.9-18          Not Applicable
                                                                               )
l Fig. A3.9-19          Not Applicable Fig._A, 3.9-20        Not Applicable Fig. A3.9-21          Not Applicable Fig. A3.9-22                                                                  l Not Applicable Fig. A3.9-24          Deletion Page (8/99)
Fig. A3.9-25          Not Applicable                                          i 3.11 3.11-14 3.11-13 (6/97)- 3.11-14 (8/99)                          l 3.Il 3.11-18      3.11-17 (9/90)- 3.11-18 (8/99)                          l 3.I1-45 -3.I1-46      3.Il-45(8/99)- 3.I1-46(8/99)                             !
Not Applicable 3.11-46a (8/99) - 3.11-46b (8/99)                        l Not Applicable        3.11- 46c (8/99) - Blank
: 3. I l 3. I 1-48
: 3. I l-47 (8/99) - 3. I 1-48 (Rev. 0) 3.Il 3.Il-52     3. I l- 51 (8/99) - 3. I l-51 a (8/99)                   !
Not Applicable        3.11-52 (Rev. 0) - Blank 3.11 3.11-58 3.Il-57 (8/99)- 3.I1-58 (9/89)                           !
3.I1-59.- 3.I1-60     3.11-59 (8/99)- 3.I1-60 (8/99)                           j 3.I1 3.I1-74
                                 ~
3.I1-73 (9/91)- 3.I1-74 (8/99) 3.11 3.Il-76     3.Il-75 (8/99)- 3.11-76 (6/97)                   _
3.11 3.1I-78 3.11-77 (2/99) - 3. I 1-78 (Rev. 0)                -
: 3. I l 3. I l-80
: 3. I l-79 (Rev. 0) - 3. I 1-80 (8/99)

Atixhment i
  .-                                                                 to U-603276 Page 9 of 29
[   3.Il 3.11-90              3.11-89 (9/91)- 3.Il-90 (8/99)
Fig 3.11-1                     Fig. 3. I l-1 (8/99)                           l
;     Fig. 3.11-26                   Fig. 3. I 1-26 (8/99)
R3. I I-27                     Fig. 3. I 1-27 (8/99)
Fig. 3.11-30                   Fig. 3. I 1-30 (8/99)
Fig. 3. I 1-32    -
Fig. 3. I 1-32 (8/99)
Fig. 3.11-33                   Fig. 3.11-33 (8/99)
                              $ Volume 67 4-ix x                     4-ix (11/97) x (6/97) 4.2 4.2-2                  4.2-1 (l 1/97) - 4.2-2 (Rev. 0) 4.4 4.4-4                  4.4-3 (6/97)- 4.4-4 (7/98) 4.6 4.6-16                4.6-15 (8/99)-4.6-16 (8/99) 4.6 4.6-18                4.6-17 (8/99) - 4.6-18 (Rev. 0) 5.2 5.2-8                  5.2-7 (6/97)- 5.2-8 (3/98) 5.2-9 -5.2-10 :                5.2-9 (Rev. 0) - 5.2 10 (7/98) 5.2-39-5.2-40                  5.2-39(9/94)- 5.2-40(3/98) 5.2 5.2               5.2-41 (3/98) - 5.2-42 (Rev. 0) 5.2 5.2-44                5.2-43 (Rev. 0) - 5.2-44 (8/99) m  5.2 5.2-50               5.2-49 (9/89)- 5.2-50 (8/99) 5.2 5.2-54                5.2-53 (6/97)- 5.2-54 (3/98)                   f Fig. 5.'2-16 S 1              Fig. 5.2-16 S 1 (8/99)                           (
l 5.3 5.3-8                  5.3-7 (9/91)- 5.3-8 (8/98)                      l 5.4 5.4              5.4-45 (4/96)- 5.4-46(l2/97) 5.4-47-5.4-48                 5.4-47 (12/97) .5.4-48 (12/97)         ,
5.4 5.4-50               5.4-49 (12/97)- 5.4-50 (l2/97) 5.4 5.4              5.4-51 (12/97)- 5.4-52 (12/97) 5.4 5.4-54               5.4-53 (8/99) - 5.4-54 (Rev. 5) 5.4 5.4-56       .       5.4-55 (Rev. 0) - 5.4-56 (l 1/97) 5.4 5.4-58               5.4-57 (11/97)- 5.4-58 (4/96) 5.4 5.4              5.4-65 (Rev. 0) - 5.4-66 (8/99) 5.4 5.4-68               5.4-67 (8/99)- 5.4-68 (8/99) 5.4 5.4-70               5.4-69 (8/99)- 5.4-70 (4/96) 5.4 5.4-72               5.4-71 (7/98)- 5.4-72 (4/96) 5.4 5.4-82               5.4-81 (Rev. 0) - 5.4-82 (12/97)
:     5.4 5.4-84               5.4-83 (2/99) - .5.4-84 (Rev. 0) 5.4 5.4-94               5.4-93 (3/98) - 5.4-94 (Rev. 0) l     5.4 5.4              5.4-95 (Rev. 0) - 5.4-96 (2/99) 5.4-97; 5.4-98                5.4-97 (2/99)- 5.4-98 (2/99) 5.4 5.4-100              5.4-99 (2/99)- 5.4-100 (2/99)
Fig. 5.4-2 S 1 (8/99)
[G\ Fig. 5.4-2-S 1 l
l
                 ~

1 Attachment 3 '
to U-603276 Page 10 of 29 Fig. 5.4-2 S 2      Fig. 5.4-2 S 2 (8/99)
Fig. 5.4-9 S 2      Fig. 5.4-9 S 2 (8/99)
F;g. 5.4-10         Fig. 5.4-10 (8/99)
Fig. 5.4-13 S 1     Fig. 5.4-13 S 1 (8/99)
Fig. 5.4-13 S 2     Fig. 5.4-13 S 2 (8/99)
Fig. 5.4-13 S 3     Fig. 5.4-13 S 3 (8/99)
Fig. 5.4-14 S 1     Fig. 5.4-14 S 1 (8/99)
Fig. 5.4-14 S 2     Fig. 5.4-14 S 2 (8/99)
Fig. 5.4-16 S 3     Fig. 5.4-16 S 3 (3/98)
Fig. 5.4-16 6 4     Fig. 5.4-16 S 4 (2/99)
Fig. 5.4-18        Fig. 5.4-18 (2/99) 6-iii iv       6-iii (Rev. 0) iv (8/99) 6-v vi          6-v (8/99) vi (8/99) 6-vii viii      6-vii (8/99) viii (8/99) 6.1 6.1 -4 6.1-3 (Rev. 0) - 6.1-4 (8/99) 6.1 6.1-6 6.1-5 (8/99) - 6.1-6 (Rev. 0) 6.1 6.1-8       6.1-7 (8/99)- 6.1-8 (9/91) 6.1 6.l-10      6.1-9 (8/99)- 6.1-10 (8/99)
: 6. I 6. I-12   6.1-11 (8/99)- 6.1-12 (6/97) 6.2 6.2-22     6.2-21 (Rev. 0) - 6.2-22 (2/99) 6.2 6.2-24 6.2-23 (2/99)- 6.2-24 (9/94) 6.2 6.2-26    6.2-25 (3/98)- 6.2-26 (3/98) 6.2 6 2-28     6.2-27 (3/98)- 6.2-28 (6/97) 6.2 6.2-56    6.2-55 (5/96)- 6.2-56 (8/99) 6.2 6.2-58    6.2-57 (8/99)- 6.2-58 (8/99) 6.2 6.2-60 6.2-59 (8/99) - 6.2-59a (8/99)
Not Applicable 6.2-59b (8/99) - 6.2-59c (8/99)
Not Applicable      6.2-60 (Rev. 0) - Blank 6.2 6.2-72 6.2-71 (7/98) - 6.2-72 (Rev. 0) 6.2 6.2-78     6.2-77 (5/97)- 6.2-78 (12/97) 6.2 6.2-86     6.2-85 (Rev. 0) - 6.2-86 (8/99) 6.2 6.2-88. 6.2-87 (8/99)- 6.2-88 (8/99) 6.2 6.2-90    6.2-89 (5/97)- 6.2-90 (8/99)
_6.2 6.2-100   6.2-99 (8/99)- 6.2-100 (5/97) 6.2-105 - 6.2-106  6.2-10) (11/95)- 6.2-106 (2/99) 6.2-107 - 6.2-108  6.2-107 (8/99) - 6.2-108 (Rev. 0) 6.2-119 - 6.2-120  6.2-119 (Rev. 0) - 6.2-120 (7/98) 6.2-121 - 6.2-122  6.2-121 (Rev. 0)- 6.2-122 (8/99) 6.2-123 - 6.2-124  6.2-123 (8/99)- 6.2-124 (8/99) 6.2-125 - 6.2-126  6.2-125 (8/99) - 6.2-126 (Rev. 0) 6.2-127- 6.2-128   6.2-127 (8/99) - 6.2-127a (8/99)
_Not Applicable     6.2-128 (9/89) - Blank                           l

p                                                         --                                    . _ .
Attachment 3 to U-603276 Page iI of 29 I
'O                6.2-129 - 6.2-l 30                6.2-129 (Rev. 0) - 6.2- 130 (8/99)
U              6.2-131 - 6.2-132                 6.2-131 (Rev. 0) - 6.2-132 (8/99) 6.2-175 - 6.2-176                 6.2-175 (8/99)- 6.2-176 (l2/96) 6 2-177 - 6.2-178             ,
6.2-177 (l2/96)- 6.2-178 (l/98) l                 6.2-179 - 6.2-180                 6.2-179 (8/99)- 6.2-180 (5/97) l                 6.2-181 - 6.2-182    .
6.2-181 (5/97)- 6.2-182 (5/95) l 6.2.I83 - 6.2-184                 6.2-183 (8/99)- 6.2-184 (8/99) l                 6.2-185 - 6.2-186-                6.2-185 (8/99)- 6.2-186 (l/98) 6.2-187 - 6.2-188                6.2-187 (8/99)- 6.2-188 (l/98)                       >
6.2 189 - 6.2-190                6.2-189 (8/99)- 6.2-190(8/99) l                 6.2-191 - 6.2-192                6.2-191 (8/94)- 6.2-192 (8/99) 6.2-193 - 6.2-194                6.2-193 (8/99)- 6.2-194 (8/99) 6.2-195 - 6.2-196                6.2-195 (Rev. 5) - 6.2-196 (Rev. 5) 6.2-197 - 6.2-198                6.2-197 (8/99) - 6.2-198 (Rev. 5)
!                 6.2-199 - 6.2-200                6.2-199 (10/94)- 6.2-200 (8/99)                        ,
6.2-201 - 6.2-202                6.2-201 (8/99) - 6.2-202 (P '99)                      ]
l 6.2-203 - 6.2-204                6.2-203 (8/99)- 6.2-204 (8/99)                       i 6.2-205 - 6.2-206                6.2-205 (8/99)- 6.2-206(8/99) l 6.2-207'- 6.2-208                6.2-207(8/99)- 6.2-208 (8/99)                         ,
l 6.2-209 - 6.2-210                6.2-209 (8/99)- 6.2-210 (8/99) 6.2-211 - 6.2-212                6.2-211 (8/99)- 6.2-212 (8/99)                        J I
6.2-213 - 6.2-214 lQ                6.2-214a - Blank 6.2-213 (8/99)- 6.2-214 (8/99)
Not Applicable 6.2-215 - 6.2-216'               6.2-215 (8/99)- 6.2-216 (8/99) 6.2-217 - 6.2-218                6.2-217 (8/99)- 6.2-218 (8/99) 6.2-219 - 6.2-220-               6.2-219 (8/99)- 6.2-220 (8/99) l 6.2-721 - 6.2-222                6.2-221 (8/99)- 6.2-222 (8/99)
!                 6.2-223 - 6.2-224                6.2-223 (8/99)- 6.2-224 (8/99) 6.2-225 - 6.2-226                6.2-225 (8/99)- 6.2-226 (8/99)
L                 6.2-227 - 6.2-227a               6.2-227 (8/99) - 6.2-227a (8/99) 6.2-227b - 6.2-228                6.2-227b (8/99) - 6.2-228 (9/90)
Fig. 6.2-55                       Fig. 6.2-55 (8/99)
M,          '       -   en 5Voluind 7?      -
Fig. 6.2-138                      Fig. 6.2-138 (8/99)
Fig. 6.2-141                      Fig. 6.2 141 (8/99)
Fig. 6.2-142                      Fig. 6.2-142 (8/99)
Fig. 6.2-143                      Fig. 6.2-143 (8/99)
(                                                                                                       f Fig. 6.2 144                      Fig. 6.2-144 (8/99)
Fig. 6.2-145                      Fig. 6.2-145 (8/99)
C: %        Fig. 6.2-147 S 1                  Fig. 6.2-147 S 1 (8/99) b L.:        .._:

Attachmcitt 3 to U-603276 Page 12 of 29 Fig. 6.2-147 S 2      Fig. 6.2-147 S 2 (8/99)
Fig. 6.2-147 3 3     Fig. 6.2-147 S 3 (8/99) 6.3 6.3-4         6.3-3 (3/97)- 6.3-4 (2/99) 6.3 6.3-6         6.3-5 (8/99)- 6.3-6 (8/99) 6.3 6.3-8         6.3-7 (8/99) - 6.3-7a (8/99)
Not Afplicable        6.3-7b (8/99) - 6.3-7c (8/99)
Not Applicable        6.3-7d (8/99) - 6.3-8 (6/97) 6.3 6.3-10        6.3-9 (Rev. 0) - 6.3-10 (8/99)
Not Applicable        6.3-10a (8/99) - 6.3-10b (8/99)
Not Applicable        6.3-10c (8/99) - Blank 6.3 6.3-12       6.3-11 (5/97)- 6.3-12 (8/99) 6.3 6.3-14       6.3-13 (8/99) - 6.3-14 (Rev. 0) 6.3-17 -6.3-18        6.3-17 (8/99)- 6.3-18 (8/99) 6.3 6.3-20      6.3-19 (5/97)- 6.3-20 (2/99) 6.3 6.3-26      6.3-25 (;2/97)- 6.3-26 (7/98) 6.3 6.3-28      6.3-27 (3/98) '6.3-28 (6/97) 6.3 6.3-32      6.3-31 (2/99)- 6.3-32 (6/97) 6.3 6.3-48       6.3-47 (7/98)- 6.3-48 (12/96) 6.3 6.3-50       6.3-49 (11/97)- 6.3-50 (6/97) 6.3 6.3-56      6.3-55 (Rev. 0) - 6.3-56 (8/99) 6.3 6.3-58      6.3-57 (8/99) - 6.3-58 (Rev. 0)
Fig. 6.3-1           Fig. 6.3-1 (8/99)
Fig. 6.3-2           Fig. 6.3-2 (8/99)
Fig. 6.3-4           Fig. 6.3-4 (8/99)
Fig. 6.3-5            Fig. 6.3-5 (8/99)
Fig. 6.3-8 S 1        Fig. 6.3-8 S 1 (8/99)
Fig. 6.3-8 S 2        Fig. 6.3-8 S 2 (8/99)
Fig. 6.3-69           Fig. 6.3-69 (7/98)
Fig. 6.3-70          Fig. 6.3-70 (7/98)
Fig. 6.3-71          Fig. 6.3-71 (7/98)
Fig. 6.3-72          Fig. 6.3-72 (7/98) 6.4 6.4-2        6.4-1 (8/99)- 6.4-2 (8/99) 6.4 6.4-4        6.4-3 (8/99)- 6.4-4 (8/99) 6.4 6.4-6        6.4-5 (8/99)- 6.4-6 (l/96) 6.4 6.4-8        6.4-7 (1/96)- 6.4-8 (8/99)                       j 6.4-9 -6.4-10        6.4-9 (8/99)- 6.4-10 (8/99)                      j Fig. 6.4-1           Fig. 6.4-1 (2/99) 6.5 6.5-26      6.5-25 (7/98) - 6.5-26 (Rev. 0) 6.5 6.5-38      6.5-37 (Rev. 0) - 6.5-38 (l/98) 6.5-4926.5-50        6.5-49 (1/98)- 6.5-50 (1/98)                     )
6' 7 6.7-4 6.7-3 (Rev. 0) - 6.7-4 (12/97) 6.7 6.7-6        6.7-5 (Rev. 0) - 6.7-6 (12/97) e--_ _ _ _ _ ~ _ - - - - - - - ~           ---

Attachmem 3 e                                                                to U-603276 Page 13 of 29 6.7 6.7-8          6.7-7 (8/99) - 6.7-8 (Rev. 0) j              6.7 6.7-10         6.7-9 (7/98) - 6.7-10 (Rev. 5)
Fig. 6.7-1             Fig. 6.7-1 (12/97) 6.8- 1                 6.8-1 (8/99)
;              7-i ii -           7-i (Rev. 0) ii (8/99) 7-xvii xviii -   -
7-xvii (Rev. 0) xviii (8/99) 7.xix xx           7-xix (8/99) xx (8/99) 7-xxi xxii         7-xxi (8/99)- 7-xxii (8/99) 7-xxiii xxiv -     7-xxiii (8/99) xxiv (8/99) 7-xxxv .7-xxxvi -      7-xxxv (1/98) xxxvi (Rev. 0) .
7-xliii xliv       7-xliii (Rev. 0) xliv (2/99) 7-xlv xlvi         7-xlv (2/99) -' 7-xlvi (Rev. 0) 7-lv Ivi :         7-IV (Rev. 0) Ivi (8/99) 7-lvii lviii       7-lvii (8/99) lviii (Rev. 0) 7-lxi lxii         7-Ixi (8/99) lxii (8/99) 7-lxiii lxiv       7-lxiii (8/99) lxiv (Rev. 0)
              '7-lxxi lxxii       7-lxxi (Rev. 0) lxxii (8/99) 7-lxxiii lxxiv     7-lxxiii (8/99) lxxiv (8/99) 7-lxxv lxxvi '     7-lxxv (8/99) lxxvi (8/99) 7-lxxii -              7-lxxxi (2/99) 7.1 7.1-2 <    .
7.1-1 (Rev. 0) - 7.1-2 (8/99) s       7.1-57.1-6             7.1-5 (8/99) - 7.1-6 (Rev. 0) 7.1 7.1-18        7.1-17 (Rev. 0) - 7.1-18 (8/99) 7.1 7.1-20'        7.1-19 (8/99)- 7.1-20 (8/99) 7.1 7.1-22        7.1-21 (8/99) - 7.1-22 (Rev. 0) 7.1 7.1-28         7.1-27 (l/96)- 7.1-28 (7/98) 7.1 7.1-64        7.1-63 (Rev. 0) - 7.1-64 (1/98)
              '7.1 7.1       7.1-80 (6/97)- 7.1-81 (8/99) 7.1 7.1-87        7.1-86 (Rev. 0) - 7.1-87 (8/99) 7.1-104'-'7.1-105      7.1-104 (8/99) - 7.1-105 (Rev. 0) 7.1-108 - 7.1-109      7.1-108 (8/99)- 7.1-109 (8/99) 7.1-110 - 7.1-111      7.1-110 (8/99) - 7.1-111 (Rev. 0) 7.1-112 - 7.1-113      7.1-112 (8/99) - 7.1-113 (Rev. 0) 7.1-120 - 7.1-121      7.1-120 (8/99) - 7.1-121 (Rev. 0)
Fig. 7.1-4             Fig: 7.1-4 (8/99) 7.2-15 -'7.2-16        7.2-15 (3/98)- 7.2-16 (8/94) 7.2 7.2       7.2-43 (Rev. 0) - 7.2-44 (8/99) 7.2 7.2-54        7.2-53 (3/98) - 7.2-54 (Rev. 0) 7.2 7.2-56        7.2-55 (2/95) 7.2-56 (3/98) l 7.2 7.2-58         7.2-57 (Rev. 0) - 7.2-58 (3/98) i.-
7.2 7.2-60'       7.2-59 (Rev. 0) - 7.2-60 (3/98) 7.2 7.2-68        7.2-67 (Rev. 0) - 7.2-68 (3/98)                -
    %_                                                                                         I e

I j
l Attachment 3        ,
to U-603276 Page 14 of 29 Fig. 7.2 10               Fig. 7.2-10 (8/99)                             -
7.3 7.3-14           7.3-13 (1/98)- 7.3-14 (2/95) 7.3 7.3-18           7.3-17 (Rev. 0) - 7.3-18 (3/98) 7.3 7.3-22           7.3-21 (9/89)- 7.3-22 (3/98) 7.3 7.3-26           7.3-25 (5/97)- 7.3-26(2/99) 7.3 7.3 28          7.3-27 (7/98)- 7.3-28 (2/99) 7.3-29 -7.3-30           7.3-29 (12/94)- 7.3-30 (8/99) 7.3 7.3-32          7.3-31 (Rev. 0) - 7.3-32 (7/98) 7.3-47-7.3-48            7.3-47 (12/97)- 7.3-48 (5/97) 7 3 7.3-64          7.3-63 (Rev. 0) - 7 3-64 (8/99) 7.3 7.3-66          7.3-65 (8/99). 7.3-66 (12/97) 7.3 7.3-70          7.3-69 (8/99)- 7.3-70 (8/99) 7.3 7.3-72          7.3-71 (8/99)- 7.3-72 (8/99) 7.3 7.3-74          7.3-73 (8/99)- 7.3-74 (8/99) 7.3 7.3-76          7.3-75 (8/99)- 7.3-76 (8/99) 7.3 7.3-82          7.3-81 (Rev. 0) - 7.3-82 (8/99)
Not Applicable          7.3-82a (8/99) - Blank 7.3 7.3-90         7.3-89 (Rev. 0) - 7.3-90 (8/99) 7.3 7.3-94         7.3-93 (8/99) - 7.3-94 (Rev. 0) 7.3 7.3-98         7.3-97 (8/99) - 7.3-98 (Rev. 0) 7.3 7.3-100        7.3-99 (8/99)- 7.3-100 (8/99) 7.3-103 - 7.3-103a       7.3-103 (9/89) - 7.3-103a (8/99) 7.3-103b - 7.3-104       7.3-103b (8/99) - 7.3-103c (8/99)
Not Applicable           7.3-104 (12/94) - Blank 7.3-189 - 7.3-190        7.3-189 (3/98)- 7.3-190 (5/95)
Fig. 7.3-4               Fig. 7.3-4 (8/99)
Volume 8 -
7.4 7.4-2            7.4-1 (6/97)- 7.4-2 (12/97) 7.4 7.4-4           7.4-3 (8/99) - 7.4-4 (Rev. 0) 7.4 7.4-6           7.4-5 (8/99)- 7.4-6 (3/98) 7.4 7.4-8           7.4-7 (8/99)- 7.4-8 (6/96) 7.4 7.4-12         7.4-11 (7/98)- 7.4-12 (8/99) 7.4 7.4-20          7.4-19(8/99)- 7.4-20 (8/99) 7.4 7.4-22         7.4-21 (8/99)- 7.4-22.(8/99)                          l 7.4 7.4-24         7.4-23 (8/99)- 7.4-24 (4/96) 7.4 7.4-26         7.4-25 (Rev. 0) - 7.4-26 (8/99) 7.4 7.4-29a        7.4-29 (2/99) - 7.4-29a (8/99) 7.5 7.5-2           7.5-1 (8/99)- 7.5-2 (8/99) 7.5 7.5-6           7.5-5 (8/99) - 7.5-6 (Rev. 5)                         l 7.5 7.5-16         7.5-15 (Rev. 0) - 7.5-16 (8/99)                 ,
1

                                                                       . . . . ~ . . - . .
;-e Attachment 3
  .+                                                     to U-603276 Page 15 of 29 7.5 7.5-18       7.5-17 (8/99)- 7.5.I8 (8/99) 7.5 7.5-20      '7.5-19 (8/99) - 7 5-20 (Rev. 0) 7.5-39-7.5-40        7.5-39 (7/98)- 7.5-40 (7/98) 7.5 7.5-42        7.5-41 (7/98)- 7.5-42 (7/98) 7.5 7.5-44      7.5-43 (7/98) - 7.5-44 (Rev. 0)
Fig. 7.5-3           Fig. 7.5-3 (8/99)-
7.6 7.6 28      7.6-27 (5/97)- 7.6-28 (1/98) i    7.6 7.6-34      7.6-33 (Rev. 0) - 7.6-34 (2/99) l    7.6 7.6-38      7.6-37 (Rev. 0) - 7.6-38 (8/99) 7.6 7.6-40      7.6-39 (8/99) - 7.6-40 (Rev. 0)
:    7.6-45-7.6-46        7.6-45 (Rev. 0) - 7.6-46 (8/99)
Not Applicable       7.6-46a (8/99) - 7.6-46b (8/99)
Not Applicable       7.6-46c (8/99) - Blank 7.6 7.6-52      7.6-51 (12/97)- 7.6-52 (12/97)                                   ,
7.6 7.6-66      7.6-65 (Rev. 0) - 7.6-66 (S/99) -
7.6 7.6-88      7.6-87 (Rev. 0) - 7.6-88 (8/99)
Not Applicable       7.6-88a (8/99) - 7.6-88b (8/99)
Not Applicable       7.6-88c (8/99) . Blank 7.6-105 - 7.6-106    7.6-105 (Rev. 0) - 7.6-106 (1/98) 7.7 7.7-2        7.7-1 (Rev. 0) - 7.7-2 (8/99) 7.7 7.7-16      7.7-15 (Rev. 0) - 7.7- 16 (8/99)
V 7.7 7.7-48      7.7-47(8/99)- 7.7-48 (l/97) 7.7 7.7-72      7.7-71 (8/99) - 7.7-72 (Rev. 0) 7.7 7.7-74      7.7-73 (2/99) - 7.7-74 (Rev. 0) 7.7-111 - 7.7-112    7.7-111 (8/99)- 7.7-112 (8/99) 7.7-113 - 7.7-114    7.7-113 (8/99)- 7.7-114 (12/94) 7.7-123 - 7.7-124    7.7-123 (3/98)- 7.7-124 (8/99) 7.7-125 - 7.7-125a   7.7-125 (8/99) - 7.7-125a (8/99)                                 ]
7.7-125b - 7.7-126   7.7-125b (8/99) - 7. 7-126 (3/97)                                l Not Applicable       7.7-130a (8/99) - 7.7-130b (8/99)
Not Applicable       7.7-130c (8/99) - 7.7-130d (8/99)
Not Applicable       7.7-130e (8/99) - Blank 7.7-149- 7.7-150-    7.7-149 (Rev. 0) - 7.7-150 (2/99)                                 i l
7.7-151 - 7.74152    7.7-151 (Rev. 0) - 7.7-152 (8/99) 7.7-153 - 7.7 154    7.7-153 (8/99) - 7.7-154 (Rev. 0)
Fig. 7.7-2           Fig. 7.7-2 (2/99) -
8-i ii           8-i (3/98) ii (Rev. 0) 8-iii iv         8-iii (2/99)- 8-iv (2/99) 8.1 8.1-4        8.1-3 (8/99)- 8.1-4 (3/98) 8.1-78.1          8.1-7 (7/98) - 8.1-8 (Rev. 0) 8.1 8.1-14      8.1-13 (Rev. 0) - 8.1-14 (7/98) 8.1 8.1     8.1-15 (7/98)- 8.1-16 (2/99) o

e Attachment 3    .
to U-603276 Page 16 of 29 8.1 8.1-18      8.1-17 (2/99)- 8. I- 17a (2/99)
Not Applicable        8.1-18 (Rev. 0) - Blank 8.1 8.1-20       8.1-19 (Rev. 0) - 8.1-20 (2/99) 8.2 8.2-6         8.2-5 (l0/94) - 8.2-6 (8/99)
Fig. 8.2-2            Fig. 8.2-2 (8/99) 8.3 8.3-2         8.3-1 (8/99)- 8.3-2 (8/99) 8.3 8.3-4        8.3-3 (8/99)- 8.3-4 (8/99) 8.3 8.3-6        8.3-5 (8/99)- 8.3-6 (8/99) 8.3 8.3-8        8.3-7 (8/99)- 8.3-8 (8/99) 8.3 8.3-10       8.3-9 (8/99)- 8.3-10 (5/97) 8.3 8.3-14      8.3-13 (Rev. 0)- 8.3-14 (7/98) 8.3 8.3-16      8.3-15 (8/99)- 8.3-16 (8/99) 8.3 8.3-20      8 3-19 (Rev. 5) - 8.3-20 (2/99) 8.3 8.3-24      8.3-23 (2/99)- 8.3-24 (9/92) 8.3 8.3-26      8.3-25 (2/99) - 8.3-25a (2/99)
Not Applicable       8.3-26 (6/97) - Blank 8.3 8.3-28       8.3-27 (12/97)- 8.3-28 (8/99) 8.3-31-8.3-32        8.3-31 (Rev. 0) -8.3-32 (8/99) 8.3 8.3-34     8.3-33 (8/99) - 8.3-33a (2/99)
Not Applicable      8.3-34 (9/89) - Blank                              l 8.3 8.3-44     8.3-43 (Rev. 0) - 8.3-44 (8/99) 8.3 8.3-46      8.3-45(8/99)-8.3-46(8/99) 8.3 8.3-48   _  8.3-47 (8/99)- 8.3-48 (8/99) 8.3-49-8.3-50       8.3-49 (8/99)- 8.3-50 (8/99) 8.3 8.3-58      8.3-57 (l/98)- 8.3-58 (2/99) 8.3 8.3-64      8.3-63 (8/99) - 8.3-64 (Rev. 0) 8.3 8.3-84      8.3-83 (8/99)- 8.3-84 (8/99) 8.3 8.3-88      8.3-87 (Rev. 0) - 8.3-88 (8/99) 8.3-88a - 8.3-88b    8.3-88a (8/99) - 8.3-88b (8/99) 8.3-88c - 8.3-88d    8.3-88c (8/99) - 8.3-88d (8/99) 8.3-88e - 8.3-88f    8.3-88e (8/99) - 8.3-88f(8/99) 8.3 8.3-90     8.3-89 (8/99) - 8.3-90 (Rev. 0) 8.3 8.3-96     8.3-95 (6/97)- 8.3-96 (8/99) 8.3 Blank      8.3-98 (8/99) - Blank                              }
Fig. 8.3-1          Fig. 8.3-1 (8/99)                                  i Fig. 8.3 2 S 2      Fig. 8.3-2 S 2 (8/98)
Fig. 8.3-7          Fig. 8.3-7 (2/98)
Fig. 8.3-9 S 1      Fi'g. 8.3-9 S 1 (8/99)
Fig. 8.3-9 S 3      Fig. 8.3-9 S 3 (8/99)
Fig. 8.3-9 S 4      Fig. 8.3-9 S 4 (8/99)
. Fig. 8.3-9 S S      Fig. 8.3-9 S 5 (8/99)
Fig. 8.3-9 S 6      Fig. 8.3-9 S 6 (8/99)                          -

Attachmetil 3 to U-603276 Page 17 of 29 Fig. 8.3 9 S 7      Fig 8.3-9 S 7 (8/99)
Fig. 8.3-9 S 8      Fig. 8.3-9 S 8 (8/99)
Fig. 8.3-9 S 10     Fig. 8.3-9 S 10 (8/99 Fig. 8.3-9 S 11     Fig. 8.3-9 S 11 (8/99)
Fig. 8.3-9 S 12     Fig. 8.3-9 S 12 (8/99)
Not Applicable . Fig. 8.3-9 S 13 (8/99)
Not Applicable      Fig. 8.3-9 S 14 (8/99)                      _
Not Applicable =    Fig. 8.3-9 S 15 (8/99)
Not Applicable     Fig. 8.3-9 S 16 (8/99)
Not Applicable     Fig. 8.3-9 S 17 (8/99) 9-i ii         9-i (6/96) ii (8/99) 9-iii iv       9-iii (8/99) iv (8/99) 9-v vi         9-v (6/96) vi (8/99) 9-vii viii     9-vii (8/99) viii (Rev. 0) 9-ix x         9-ix (2/99) x (8/99) 9-xi xii       9 xi (8/99) xii (8/99) 9-xiii xiv     9-xiii (8/99) xiv (Rev. 0) 9.1 9.1-14    9.1-13 (9/92)- 9.1-14 (8/99) 9.1 9.1-16    9.1-15 (8/99)- 9.1-16 (12/94) 9.1-'17 - 9.1-18   9.1-i7 (8/99)- 9.1-18 (8/99)
(3   9.1-18a - Blank    9.1-18a (8/99) - Blank Iv / 9.1 9.1   9.1-19 (8/99)- 9.1-20 (12/96) 9.2 9.2-2      9.2-1 (8/99)- 9.2-2 (8/99) 9.'2 9.2-4     9.2-3 (8/99) - 9.2-3a (8/99)
Not Applicable     9.2-4 (8/99) - Blank 9.2 9.2-6      9.2-5 (3/98) - 9.2-6 (Rev. 0) 9.2 9.2-8      9.2-7 (8/99) - 9.2-8 (Rev. 0)                      J 9.2 9.2-22    9.2-21 (8/99) - 9.2-22 (Rev. 0)                    l 9.2 9.2-24    9.2-23 (8/99) - 9.2-24 (Rev. 0) 9.2 9.2-30    9.2-29 (8/99)- 9.2-30 (8/99) 9.2 9.2-34    9.2-33 (8/99)- 9.2-34 (8/99)                       i 9.2 9.2-36    9.2-35 (8/99) - 9.2-36 (Rev. 0) 9.2 9.2-38    9.2-37 (8/99)- 9.2-38 (5/96) 9.2 9.2-40    9.2-39 (8/99) - 9.2-40 (Rev. 0) 9.2 9.2-42     9.2-41 (8/99) - 9.2-42 (Rev. 0) 9.2 9.2-44    9.2-43 (Rev. 0) - 9.2-44 (8/99) 9.2 9.2-48     9.2-47 (Rev. 0) - 9.2-48 (8/99)
Fig. 9.21 S 1      Fig. 9.2-1 S 1 (8/99)
Fig. 9.2-1 S 2     Fig. 9.2-1 S 2 (8/99)
Fig. 9.2-2 S 1     Fig. 9.2-2 S 1 (8/99)
Fig. 9.2-2 S 2     Fig. 9.2-2 S 2 (8/99)
Fig. 9.2-2 S 3     Fig. 9.2-2 S 3 (8/99)                               j p

Attachment 3    ~
to U-603276 Page 18 of 29 Fig. 9.2-3 S 3                            l Fig 9.2-3 S 3 (8/99)
Fig. 9.2-11 S 1                             Fig. 9.2-11 S 1 (12/97)
Fig. 9.2-11 S 5                             Fig. 9.2-11 S 5 (8/99)
Fig. 9.2-11 S 6                             Fig. 9.2-11 S 6 (8/99)
Fig. 9.2-12 S 1                             Fig. 9.2-12 S I (8/99)
Fig. 9.2-12 S 2                             Fig. 9.2-12 S 2 (3/98)
Fig. 9.2-16 S 1                      ,
Fig. 9.2-16 S 1 (12/97)
                                        - Volume 9 i 9.3 9.3-2                               9.3-1 (8/99)- 9.3-2 (8/99) 9.3 9.3-4                               9.3-3 (8/99)- 9.3-4 (8/99) 9.3 9.3-6                               9.3-5 (Rev. 0) - 9.3-6 (8/99) 9.3 9.3-8                              9.3-7 (8/99)- 9.3-8 (6/96) 9.3 9.3-12                            9.3-11 (8/99)- 9.3-12 (8/99) 9.3 9.3-14                            9.3-13 (8/99)- 9.3-14 (8/99)                 ~
9.3 9.3-16                            9.3-15 (8/99)- 9.3-16 (2/99) 9.3 9.3-26                            9.3-25 (8/99) - 9.3-26 (Rev. 5) 9.3 9.3-30                            9.3-29 (l I/97) - 9.3-30 (Rev. 0) 9.3 9.3-32                             9.3-3i (2/99)- 9.3-32 (8/99)                     ,
l 9.3 9.3-34                             9.3-33 (11/97)- 9.3-34 (11/97) 9.3 9.3-36                             9.3-35 (11/97)- 9.3-36 (l1/97) 9.3 9.3-38                             9.3-37 (11/97)- 9.3-38 (11/97) 9.3 9.3-40                              9.3-39 (11/97)- 9.3-40 (11/97)                   l Fig. 9.3-1 S 1                              Fig. 9.3-1 S 1 (12/97)
Fig. 9.3-1 S 8                              Fig. 9.3-1 S 8 (3/98)
Fig. 9.3-1 S 9                              Fig. 9.3-1 S 9 (8/99)
Fig. 9.3-2 S 4                              Fig. 9.3-2 S 4 (8/99)                            l Fig. 9.3-2 S S                              Fig. 9.3-2 S 5 (2/99)
Fig. 9.3-2 S 6                             Fig. 9.3-2 S 6 (7/98)
Fig. 9.3-2 S ?                             Fig. 9.3-2 S 7 (8/99)
Fig. 9.3-2 S 10 f
Fig. 9.3-2 S 10 (8/99)                            j Fig. 9.3-2 S 13                            Fig. 9.3-2 S 13 (8/99)
Fig. 9.3-2 S 14                            Fig. 9.3-2 S 14 (8/99)                              ,
_Fjg. 9.3-2 S 16                            Fig. 9.3-2 S 16 (8/99)
Fig. 9.3-2 S 21                            Fig. 9.3-2 S 21 (8/99)                              l Fig. 9.3-2 S 22                            Fig. 9.3-2 S 22 (8/99)                              !
Fig. 9.3-3 S 1 through Fig. 9.3-3 S8       Fig. 9.3-3 S 1 (8/99)
Fig. 9.3-4 S 5                             Fig. 9.3-4 S 5 (8/99)
Fig. 9.3'-4 S 6 Fig. 9.3-4 S 6 (8/99)
Fig. 9.3-4 S 9                             Fig. 9.3-4 S 9 (8/99) 9.41-9.4-2                                 9.4-1 (8/99) - Blank

l Attachment 3 to U-603276 Page 19 of 29
       )   Not Applicable      9.4-Ia (8/99) - 9.4-2 (8/99) 9.4 9.4-6       9.4-5 (Rev. 0) - 9.4-6 (8/99)
Not Applicable      9.4-6a (8/99) - Blank 9.4 9.4-10     9.4-9 (8/99) - Blank Not Applicable     9.4-9a (8/99) - 9.4-10 (l/97) 9.4 9.4-22    9.4-21 (9/89)- 9.4-22 (8/99) 9.4 9.4-24    9.4-23 (8/99) - Blank Not Applicable     9.4-23a (8/99) - 9.4-24 (9/89) 9.4 9.4-30    9.4-29 (8/99)- 9.4-30 (9/92) 9.4 9.4-38    9.4-37 (8/99)- 9.4-38 (8/99) 9.4-39 -9.4-40     9.4-39 (2/98) - 9.4-40 (Rev. 0) 9.4 9.4-48     9.4-47 (8/99) - 9.4-48 (Rev. 0) 9.4 9.4-60    9.4-59 (3/98)- 9.4-60 (2/96)                                 !
9.4 9.4 66    9.4-65 (2/99)- 9.4-66 (2/99) 9.4 9.4-72    9.4-71 (9/90)- 9.4-72 (3/98) 9.4 9.4-74    9.4-73 (3/98)- 9.4-74 (6/96)
Fig. 9.4-1 S 4     Fig. 9.4-1 S 4 (2/99)
Fig. 9.4-5         Fig. 9.4-5 (8/99)
Fig. 9.4-14    . Fig. 9.4-14 (8/99)
Fig. 9.4-18 S 5    Fig. 9.4-18 S 5 (8/99)
Fig. 9.4-21 S 14   Fig. 9.4-21 S 14 (2/99)
{~])
L      9.5 9.5-2      9.5-1 (3/98)- 9.5-2 (8/98) 9.5-2a - Blank     9.5-2a (8/99) - Blank 9.5 9.5-4      9.5-3 (8/99)- 9.5-4 (3/97) 9.5 9.5-6      9.5-5 (8/99)- 9.5-6 (8/99) 9.5 9.5-8      9.5-7 (8/99)- 9.5-8 (3/98)                                   l 9.5 9.5-10     9.5-9 (Rev. 0) - 9.5-10 (8/99) 9.5 9.5-12    9.5-1I (8/99)- 9.5-12 (8/95) 9.5-13 -9.5-14     9.5-13 (8/99) - Blank Not Applicable     9.5-13a (8/99) - 9.5-14 (8/99) 9.5 9.5-16     9.5-15 (8/99) - Blank Not Applicable     9.5-15a (8/99) - 9.5-16 (8/99) 9.5 9.5-18    9.5-17 (8/99)- 9.5-18 (8/99) 9.5 9.5-20    9.5-19 (8/99)- 9.5-20 (8/99) 9.5 9.5-22    9.5-21 (11/97) - 9.5-22 (Rev. 0) 9.5 9.5-28    9.5-27 (8/99)- 9 5-28 (8/99) 9.5-29 -9.5-30     9.5-29 (8/99) - 9.5-30 (Rev. 0) 9.5-45-9.5-46      9.5-45 (2/99) - 9.5-46 (Rev. 5) 9.5-47 -9.5-48     9.5-47 (Rev. 0) - 9.5-48 (3/98)
Fig. 9.S-1 S 1     Fig. 9.5-1 S 1 (8/99)
Fig. 9.5-1 S 2     Fig. 9.5-1 S 2 (8/99)
         , Fig. 9.5-1 S 5     Fig. 9.5-1 S 5 (8/99)

                                                                                - ~ .
Attachment 3        -
to U-6P3276 Pap 2i)of 29 Fig. 9.5-1 S 8                 Fig. 9.5-1 S 8 (3/98)
Fig. 9.5-1 S 9                 Fig. 9.5-1 S 9 (8/99)
Fig. 9.5-4 S 1                 Fig. 9.5-4 S 1 (8/99)
Fig. 9.5-4 S 2                 Fig. 9.5-4 S 2 (8/99) 10-iii iv                 10-iii (8/99) iv (Rev. 0) 10.1 10.1-2                10.1-1 (8/99) - 10.1-2 (Rev. 0)
Fig.10.2-5                    Fig.10.2-5 (12/97) 10.3 10.3-2                10.3-1 (Rev. 0)- 10.3-2 (2/99)
Fig.10.3-1 S 2                Fig.10.3-1 S 2 (2/99) 10.4 10.4-10 10.4-9 (l/96)- 10.4-10 (8/99) 10.4 10.4-14 10.4-13 (12/97)- 10.4-14 (5/95) 10.4 10.4-16 10.4-15 (5/97)- 10.4-16 (8/99) 10.4 10.4-18 10.4-17 (8/99)- 10.4-18 (8/99) 10.4 19- 10.4-20              10.4-19 (8/99)- 10.4-20 (8/99)
_10.4 10.4             10.4-21 (8/99)- 10.4-22 (8/99) 10.4-23 -10.4-24               10.4-23 (8/99)- 10.4-24 (8/99) 10.4 10.4-26 10.4-25 (8/99) - 10.4-26 (Rev. 0)
Fig.10.4 1                     Fig.10.4-1 (8/99)
Fig.10.4-3                    Fig.10.4-3 (8/99)
Fig.10.4-5 S 1                Fig.10.4-5 S 1 (8/99)
Fig 10.4-6                    Fig.10.4-6 () 2/97)
Fig.10.4-7                    Fig.10.4-7 (8/99)
Fig.10.4-8 S 1                Fig.10.4-8 S 1 (2/99) k Voliime 10"-
Fig. I1.2-2 S 19               Fig. I1.2-2 S 19 (3/98)
Fig. I1.2-2 S 20               Fig. I1.2-2 S 20 (3/98) 11.3 11.3-32             11.3-31 (2/99) - 11.3-32 (Rev. 0) 11.3 Blank               11.3-41 (2/99) - Blank 11.4 11.4-4 11.4-3 (9/91)- 11.4-4 (2/99) 11.4 11.4-6 11.4-5 (9/92)- 11.4-6 (2/99) 11.4 11.4-10 11.4-9 (4/97)- 11.4-10 (2/99) 11.4 11.4-16 11.4-15 (Rev. 0) - 11.4-16 (8/99)
Fig. I 1.4-1                  Fig. I1.4-1 (3/98)
Fig. I1.4-3 S 2               Fig. I1.4-3 S 2 (8/98)
Fig. 11.4-3 S 3               Fig. I1.4-3 S 3 (8/98)
Fig. I1.4-3 S 4               Fig. I1.4-3 S 4 (7/98) 11.5-7 _11.5-8                11.5-7 (7/98)- 11.5-8 (7/98) 12-ix - i2-x                  12-ix (3/98) x (Rev. 0) 12.1 12.1-2 12.1-1 (3/98)- 12.1-2 (3/97)
Fig.12.1-1                    Fig.12.1-1 (3/98)                              /
s

s______.____..     .
Attachmciit 3 to U-603276 Page 21 of 29
  'f        12.2 12.2-8          12.2-7 (8/99)- 12.2-8 (l/95) 12.2 12.2-34 12.2-33 (8/99)- 12.2-34 (8/99) 12.3 1 - 12.3-2          12.3-1 (9/95)- 12.3-2 (2/99)                      _
12.3 12.3-4          12.3-3 (3/98)- 12.3-4 (3/98) 12.3-5 12.3-6            12.3-5 (3/98)- 12.3-6 (6/97) 12.3 12.3-8     -
12.3-7 (1/96)- 12.3-8 (3/98) 12.3 12.3-10          12.3-9 (3/98)- 12.3-10 (3/98) 12.3 12.3-12         12.3-11 (3/97)- 12.3-12 (8/99) 12.3 12.3-16        12.3-15 (Rev. 0)- 12.3-16 (1/98)                             q 12.3 17 - 12.3-18         12.3-17 (Rev. 5)- 12.3-18 (8/99) 12.3 12.3-20         12.3-19 (8/99)- 12.3-20 (8/99) 12.3 12.3-22         12.3-21 (1/98)- 12.3-22 (8/99)                               1 12.3 12.3-24         12.3-23 (8/99)- 12.3-24 (3/98) 12.3 12.3-26         12.3-25 (8/99)- 12.3-26 (l/97)                                I 12.3 12.3-38         12.3-37 (7/98)- 12.3-38 (l/97)
Fig.12.3-4               Fig.12.3-4 (8/99)
Fig.12.3-13              Fig.12.3-13 (8/98)
Fig.,12.3-23             Fig 12.3-23 (8/99)           '
Fig.12.3-24              Fig.12.3-24 (8/99)
Fig.12.3-38              Fig.12.3-38 (8/98) i
_F_ig.12.3-42             Fig.12.3-42 (8/99) i  Fig.12.3-44              Fig.12.3-44 (8/99)
Fig.12.3-51              Fig.12.3-51 (8/99)
Fig.12.3-63              Fig.12.3-63 (8/99) 12.5 12.5-2          12.5-1 (2/99)- 12.5-2 (2/99) 12.5-3 '12.5          12.5-3 (3/97)- 12.5-4 (8/98) 12.5 12.5         12.5-7 (3/97)- 12.5-8 (11/97) 12.5 12.5-10         12.5-9 (2/96)- 12.5-10 (3/98)
;          12.5 12.5-12        12.5-11 (3/98)- 12.5-12 (3/97)
Fig.12.5-1               Fig.12.5-1 (8/98) 13-i ii-            13-i (8/99)- 13-ii (2/99) 13-ii! iv           13-iii (2/99) iv (3/98) 13.1 13.1-2          13.1-1 (3/98)- 13.1-2 (7/98) 13.1 13.1-4'         13.1-3 (7/98)- 13.1-4 (3/98)
          '13.1 13.1-6          13.1-5 (2/99)- 13.1-6 (8/99) 13.1 13.1-8          13.1-7 (2/99)- 13.1-8 (2/99) 13.1 13 1-10         13.1-9 (2/99)- 13.1-10 (2/99) 13.1 13.1-12        13.1-11 (8/99)- 13.1-12 (8/99) 13.1 13.1-14        13.1-13 (8/99)- 13.1-14 (8/99) 13.1 13.1 16:       13.1-15 (8/99)- 13.1-16 (8/99) 13.1 13.1-18 ~      13.1-17 (8/99)- 13.1-18 (8/99)
[3
   %/
13.1 13.1-20'      '13.1 19 (8/99)- 13.1-20 (2/99)                   l
                                 +

Attachment 3        ,
to U-603276 Page 22 of 29 13.1 13.1-22                   13.1-21 (8/99)- 13.1-22 (2/99) 13.1 13.1-24                   13.1-23 (2/99)- 13.1-24 (2/99)                         i 13.1 13.1-26                    13.1-25 (3/98)- 13.1-26 (8/99)                   ,
Fig.13.1 -1                         Fig.13.1-3 (8/99)
Fig.13.1-1 A                        Deletion Page (3/98)
Fig 13.1-1B (Deletion Page)         Not Applicable Fig.13.1-2 (Deletion Page)          Not Applicable Fig.13.1-3                          Not Applicable Fig.13.1-4 (Deletion Page)          Not Applicable 13.2 13.2-2                     13.2-1 (8/99)- 13.2-2 (8/99) 13.2 13.2-4                     13.2-3 (8/98)- 13.2-4 (8/98)
                                                                                             ]
13.2 13.2-6                     13.2-5 (3/98)- 13.2-6 (7/98)                            j 13.2 13.2-8                     13.2-7 (Rev. 5) - 13.2-8 (8/98) 13.2 13.2-10                    13.2-9 (3/97)- 13.2-10 (2/98) 13.2 13.2-14                   13.2-13 (7/98)- 13.2-14 (12/96) 13.4 13.4-2                     13.4-1 (4/97)- 13.4-2 (8/99) 13.4 13.4-4                     13.4-3 (8/99) - Blank Not Applicable                       13.4-3a (8/99)- 13.4-4 (8/99) 13.4 13.4-6                     13.4-5 (2/98)- 13.4-6 (8/99) 13.4 Blank                      13.4-7 (8/99) - Blank 13.5 13.5 2                    13.5-1 (6/97)- 13.5-2 (8/99)                     .
13.5 13.5-4                    13.5-3 (12/97)- 13.5-4 (2/99) 13.5 13.5-6                    13.5-5 (2/99)- 13.5-6 (6/97) 13.5 13.5-8                    13.5-7 (7/98)- 13.5-8 (3/98) 13.5 13.5-12                   13.5-11 (Rev. 0) - 13.5-12 (2/99) 13.5 13.5-16                   13.5-15 (9/91)- 13.5-16 (2/99) 14.2 14.2-4                     14.2-3 (Rev. 0) - 14.2-4 (3/98) 14.2-203 - 14.2-204                 14.2-203 (8/99) - 14.2-204 (Rev. 0)
Voltime' 11:
15-xve 15-xvi                       15-xv (Rev. 0) xvi (2/99)
: 15. xvii xviii                 15-xvii (2/99) xviii (Rev. 0) 15-xxiii xxiv                  15-xxiii (Rev. 0) xxiv (8/99) 15.0 15.0-8                    15.0-7 (6/97)- 15.0-8 (7/98)                            '
15.0 15.0-10                   15.0-9_(6/97)- 15.0-10 (3/98) 15.0 15.0-12                  15.0-11 (6/97)- 15.0-12 (7/98) 15.0 15.0-14                  15.0-13 (6/97)- 15.0-14 (8/99) 15.0 15.0-16                  15.0-15 (8/99)- 15.0-16 (6/97)                          I
          ~
15.0 15.0-18                  15.0-17 (7/98)- 15.0-18 (5/95) 15.1 15.1-2                    15.1-1 (6/97) - 15.1-2 (l !/9 /)                         >
15.1 15.1-8                    15.1-7 (lI/97)- 15.1-8 (7/98)                   -
l
.                                                                                            i

Attachment 3 a
to G 603276 Pagc 23 of 29 15.2 15.2                      15.2-3 (6/97)- 15.2-4 (lI/97) 15.2 15.2-8                      15.2-7 (Rev. 0) - 15.2-8 (7/98) 15.2 15.2-26                    15.2-25 (8/99)- 15.2-26 (4/96) 15.2 15.2                   15.2-27 (8/99)- 15.2-28 (8/99) 15.2 15.2-30                    15.2-29 (8/99)- 15.2-30 (4/96) 15.2 15.2-38                     15.2-37 (7/98) - 15.2-38 (Rev. 0) 15.2-39-15.2-40                      15.2-39 (Rev. 0) - 15.2-40 (l 1/97) 15.2-41 -15.2-42                     15.2-41 (7/98)- 15.2-42 (Rev. 0) 15.2 15.2-64                     15.2-63 (8/99)- 15.2-64 (8/99)
           -15.2 15.2                     15.2-65 (8/99)- 15.2-66 (8/99) 15.2-69                               15.2-69 (8/99)
Fig.15.2.6-1                         Fig.15.2.6-1 (8/99)
Fig.15.2.6-2                         Fig.15.2.6-2 (8/99)
Fig.15.2.9-2                         Fig.15.2.9-2 (7/98) -
15.3 15.3-2                       15.3-1 (Rev. 0) - 15.3-2 (7/98) 15.6 15.6-21    _                 15.6-20 (Rev. 0) - 15.6-21 (8/99) 15.6 15.6-23                    15.6-22 (8/99) - Blank Not Applicable                       15.6-22a (8/99) - 15.6-23 (6/96) 15.6 15.6-45                    15.6-44 (8/99)- 15.6-45 (Rev. 0) 15.7 15.7-16                    15.7-15 (9/92)- 15.7-16 (3/98) p       15.7 21~- 15.7-22                    15.7-21 (Rev. 0) - 15.7-22 (7/98) 15A.6 15A.6-8                    15A.6-7 (Rev. 0) - 15 A.6-8 (2/99) 15A.6 15A.6-20                  15A.6-19 (2/99) - 15 A.6-20 (Rev. 0) 15A.6 15A.6-22                  15A.6-21 (Rev. 0)- 15A.6 22 (2/99) 15A.6 15A_6-28                  15A.6-27 (6/96) - 15A.6-28 (2/99)
           -15B.7 Blank                      I5B.7-1 (Rev. 0) - 15B.7-2 (8/99)
Fig.15D.5.2-1 S 3                   Fig.1SD.5.2-1 S 3 (8/99) 15D.6 15D.6-2 15D.6-1 (6/97) - 14D.6-2 (8/99)
A A-6                            A-5 (7/98) - A'-6 (6/97)
A A-12                          A-11 (Rev. 0)- A-12 (8/99)
A A-14                          A-13 (12/96)- A-14 (8/99)
               <      ~           - > ,
Woldme 12F          '
                                                                            '&             ,      3R D-i - D-ii                           D-i (3/98)- D-ii (Rev. 0)
D D-4                            D-3 (3/98) - D-4 (Rev. 0)
D D-6                            D-5 (3/98)- D-6 (12/94)
D D-10                          D-9 (Rev. 0) - D-10 (2/99)
D D-14                          D-13 (Rev. 0) - D-14 (2/99)
D-l' - D-16                          D-15 (2/99)- D-16 (2/99)
  .-       D D-18                          D-17 (9/89)- D- 8 (3/98) b V
D D-20                          D-19 (3/98)- D-20 (Rev. 0)

Attaciunent 3    =
i lo U-603276 Page 24 of 29 D D-22          D-21 (Rev. 0) - D-22 (2/99)
D D-24          D-23 (2/99) - D-24 (Rev. 0)
D D-30         D-29 (8/99) - D-30 (Rev. 0)
D D-32         D-31 (8/99) - I)-32 (9/89)
D D-50         D-49 (8/99) - D-50 (Rev. 5)
D D-56         D-55 (8/99) - D-56 (Rev. 0)
D D-72         D-71 (8/99) - D-72 (8/99)
D D-78         D-77 (7/98) - D-78 (2/99)
D-111 - D-112       D-111 (Rev. 0) - D-112 (2/99)
E-i -E-ii           E-i (7/98) - E-ii (8/99)
E3.1 E3.1-4     E3.1-3 (2/98) - E3.1-4 (5/97)
E3.1 E3.1-6     E3.1-5 (8/99) - E3.1-6 (8/99)
E3.1 E3.1-8     E3 1-7 (8/99) - E3.1-8 (8/99)
E3.1 E3.1- 10   E3.1-9 (8/99) - E3.1-10 (12/97)
E3.2-i - E3.2-ii    E3.2-i (2/99) - E3.2-ii (8/99)
E3.2 E3.2-2    E3.2-1 (8/99) - E3.2-2 (Rev. 5)
E3.2 E3.2-4    E3.2-3 (2/99)- E3.2 4 (2/99)
E3.2 E3.2-6    E3.2-5 (2/99) - E3.2-6 (Rev. 5)
E3.2 E3.2-8    E3.2-7 (2/99) - E3.2-8 (2/99)
E3.2 E3.2-10   E3.2-9 (2/99) - E3.2-10 (2/99)                      ,
E3.2 E3.2-12 E3.2-11 (2/99) - E3.2-12 (2/99)
E3.2 E3.2-14  E3.2-13 (2/99) - E3.2-14 (2/99)
E3.2 E3.2-16   E3.2-15 (2/99) - E3.2-16 (2/99)
E3.2 E3.2-18   E3.2 17 (Rev. 5) - E3.2-18 (2/99)
E3.2 E3.2-20   E3.2-19 (2/99) - E3.2-20 (2/99)
E3.2 E3.2-22   E3.2-21 (2/99) - E3.2-22 (2/99)
E3.2 E3.2-24   E3.2-23 (2/99) - E3.2-24 (2/99)
E3.2 E3.2-26   E3.2-25 (Rev. 5) - E3.2-26 (2/99)
E3.2 E3.2-28  E3.2-27 (2/99) - E3.2-28 (2/99)
E3.2 E3.2-30  E3.2-29 (2/99) - E3.2-30 (2/99)
E3.2 Blank    E3.2-31 (8/99) - E3.2-32 (8/99)
E3.4-i - E3.4-ii   E3.4-i (2/99) - E3.4-ii (2/99)
E3.4 E3.4-2    E3.4-1 (Rev. 5) - E3.4-2 (2/99)
Not Applicable     E3.4-2a (2/99) - Blank E3.4 '7 - E3.4-8   E3.4-7 (8/99) - E3.4-8 (8/99)
E3.4 - - E3 4-10   E3.4-9 (2/99) - E3.4-10 (2/99)
E3.4 E3.4-24  E3.4-23 (6/97) - E3.4-24 (3/98)
E3.4 E3.4-26  E3.4-25 (2/99) - E3.4-2t> (5/97)
E3.4 E3.4-30 E3.4-29 (2/99) - E3.4-30 (2/99) l Ed5 E3.5-2     E3.5-1 (2/99) - E3.5-2 (Rev. 5)
E3.5 E3.5-4'   E3.5-3 (2/99)- E3.5-4 (2/99)                           '
E3.5 E3.5-6    E3.5-5 (Rev. 5) - E3.5-6 (2/99)                (w      l 1

      ,                                                                     i Anaclimen 3
      .                                                       to U-603276 Page 25 of 29
.         E3.5 E3.5-8        E3.5-7 (8/99) - E3.5-8 (Rev. 5)
        ^ E3.5 E3.5 10       E3.5-9 (2/99) - E3.5- 10 (8/99)
E3.5 E3.5-12      E3.5-11 (2/99) - E3.5-12 (2/99)
E3.5 E3.5-14      E3.5-13 (8/99) - E3.5-14 (8/99)
E3.5 E3.5-16      E3.5-15 (2/99) - E3.5-16 (Rev. 5)
E4.1 E4.1-4     -
E4.1-3 (8/99) - E4.1A(Rev.0)
E4.1 E4.1-8        E4.1-7 (Rev. 0) - E. 1-8 (8/98)
E4.1 E4.1-10       E4.1-9 (8/98) - E4.1-10 (8/95)
E4.1 E4.1-28      E4.1-27 (7/98) - E4.1-28 (8/99)
I E4.1 E4.1-30      E4.1-29 (Rev. 0) - E4.1-30 (2/99)
E4.1 E4.1-32      E4.1-31 (Rev. 0) - E4.1-32 (8/99)
E4.1 E4.1-36      E4.1-35 (8/98) - E4.1-36 (8/99)
E4.1 E4.1-38 '    E4.1-37 (8/99) - E4.1-38 (Rev. 0)
E4.1 E4.1    E4.1-45 (8/99) - E4.1-46 (8/99)
E4.1 E4.1-54     E4.1-53 (8/95) - E4.1-54 (2/99)
Fig. FP-2a            Fig. FP-2a (8/99)
Fig. FP-2b            Fig. FP-2b (8/99)
Fig. FP-3a '          Fig. FP-3a (12/97)
Fig. FP-3b            Fig. FP-3b (12/97)
Fig. FP-9a            Fig. FP-9a (2/99)      _
p-  Fig. FP-10b           Fig. FP-10b (8/09' Fig. FP-11b (8/99;
.V        Fig, FP-11b Fig. FP-12a           Fig. FP-12a (8/99)
Fig. FP-12b           Fig. FP-12b (8/99)                          !
Fig. FP-13a           Fig. FP-13a (2/99)
Fig. FP-14a           Fig. FP-14a (2/99)
Fig. FP-14b           Fig. FP-14b (2/99)                           !
Fig. FP-16a           Fig. FP-16a (2/99)                           )
Fig. FP-16b           Fig. FP-16b (2/99)
Fig. FP-19a           Fig. FP-19a (8/98)
Fig. FP-19b           Fig. FP-19b (8/98)
Fig. FP-20a           Fig. FP-20a (8/99)
Fig. FP-20b           Fig. FP-20b (8/99)                          ;
Fig. FP-24b           Fig. FP-24b (2/99)
Fig. FP-25a           Fig. FP-25a (8/99)
Fig. FP-25b           Fig. FP-25b (8/99)
Fig.- FP-27a          Fig. FP-27a ('8/99)
Fig. FP-27b -         Fig. FP-27b (8/99)
Fig. FP-28a           Fig FP-28a (8/99)      _
Fig. FP=28b           Fig FP.28b(8/99)
Fig. FP-31a           Fig. FP-3Ia (8/99)
Fig. FP-31b           Fig. FP-31b (8/99)

e Attachment 3       ,
to U-603276 q
Page 26 of 29        i Fig FP-34a Figure 2 - Cable Tray Fig. FP-34a (2/99)                               G,!
Figure 2 - Cable Tray (8/99)
Figure 3 - Cable Tray       Figure 3 - Cable Tray (12/97)
Figure 10 - Cable Tray      Figure 10 - Cable Tray (8/99)
Figure 12 - Cable Tray      Figure 12 - Cable Tray (2/99)
Figure 14 - Cable Tray      Figure 14 - Cable Tray (2/99)
Figure 17 - Cable Tray      Figure 17 - Cable Tray (8/98)
Figure 18 - Cable Tray      Figure 18 - Cable Tray (8/99)
Figure 22 - Cable Tray      Figure 22 - Cable Tray (8/99)
F-i - F-ii F-i (8/99) - F-ii (8/99)
F-iii - F-iv F-iii (8/99) - F-iv (8/99)
F-v - F-vi F-v (8/99) - F-vi (8/99)
F-vii - F-viii F-vii (8/99) - F-viii (8/99)
F-ix - F-x F-ix (8/99) - F-x (Rev. 0)
F-xvii - F-xviii F-xvii (Rev. 0) - F-xviii (8/99)
Fl.1 Fl 1-2 Fl .1-1 (8/99) - Fl .1-2 (8/99)
F1.1 Fl .1-4 Fl.1-3 (8/99) - Fl.1-4 (8/99)
Fl .1 Fl .1-6 Fl.1-5 (8/99) - Fl.1-6 (8/99)
Fl.1 Fl .1-8            Fl .1-7 (8/99) - Blank .
Not Applicable Fl.1-7a (8/99) - Fl.1-8 (9/91)
Fl .1 Fl .1-10 Fl.1-9 (8/99) - Fl.1-10 (8/99)                        j Fl .1 F.1.1-12 Fl.1-11 (8/99) - Fl .1-12 (8/99)
F1.1 F1.1-14 Fl.1-13 (8/99) - F1.1-14 (8/99)
Fl.1 Fl.1-16 11.1-15 (8/99) - F1 1-16 (8/99)                   .
F1.1 Fl.1-18 Fl .1-17 (8/99) - Fl .1-18 (8/99)
Fl.1 F1.1-20 F1.1-19 (8/99) - Fl .1-20 (8/99)                       l Fl.1 F1.1-22 Fl.1-21 (8/99) - F1.1-22 (8/99)                        {
Fl.1 Fl.1-24 Fl.1-23 (8/99) - F1.1-24 (8/99)                        {
Fl .1 Fl .1-26 Fl.1-25 (8/99) - Fl .1-26 (8/99)                       l Fl .1 Fl .1-28 Fl.1-27 (8/99) - Fl .1-23 (8/99)
F1.1 Fl.1-30 Fl.1-29 (8/99) - Fl .1-30 (8/99)                       1 Fl.1 Blank .                                                                  '
Fl.1-31 (8/99) - Fl.1-32 (8/99)
Not Applicable Fl.1-33 (8/99) - Fl.1-34 (8/99)
_ Fig. Fl.8-1 Notes         Fig. Fl.8-1 Notes (8/99)
Fig. F1.8-2 Notes          Fig. Fl .8-2 Notes (8/99)
Fig. Fl.8-3 Notes          Fig. Fl.8-3 Notes (8/99)                                l Fig. F18-4 Notes           Fig. Fl .8-4 Notes (8/99)
Fig. F1.8-5 Notes          Fig. Fl.8-5 Notes (8/99) l Fig. Fl.8-6 Notes          Fig. Fl.8-6 Notes (8/99)
Fig. Fl.'8-7 Notes         Fig. Fl .8-7 Notes (8/99)
F2.1 Blank             F2.1-1 (8/99) - Blank F3.1 F3.1-2 F3.1-1 (8/99) - F3.1-2 (8/99)

                                                                         .- . ~ . . . . . . . . .
Attachincnt 3 to U-603276 Page 27 of 29
,/ 7 V       F3.1 F3.1-4 F3.1 F3.1-6 F3.1-3 (8/99) - F3.1-4 (8/99)
F3.1-5 (8/99) - F3.1-6 (8/99)
Not Applicable      F3 1-6a (8/99) - F3.1-6b (8/99)
Not Applicable      F3.1-6c (8/99) - Blank F3.2 F3.2-2     F3.2-1 (2/99) - F3.2-2 (2/99)
F3.2 F3.2-4     F3.2-3 (8/99) - F3.2-4 (8/99)
F3.2 F3 2-6     "3.2-5 (8/99) - F3.2-6 (8/99)
F3.3 F3.3-2     .-3.3-1 (8/99) - F3.3-2 (8/99)
F3.3 3 - F3.3-4     F3 3-3 (8/99) - F3.3-4 (8/99)
F3.3 F3.3-6     F3.3-5 (8/99) - F3.3-6 (8/99)
F3.3 F3.3-8     F3.3-7 (8/99) - F3.3-8 (8/99)
F3.3 F3.3-10    F3.3-9 (8/99) - F3.3-10 (8/99)
Not Applicable      F3.3-10a (8/99) - F3.3-10b (8/99)
Not Applicable      F3.3-10c (8/99) - Blank F3.3 F3.3-12   F3 3-11 (Rev. 0) - F3.3-12 (8/99)
F3.3 F3.3-14   F3.3-13 (Rev. 0) - F3.3-14 (8/99)
F3.4 F3.4-2     F3.4-1 (8/99) - F3 4-2 (8/99)
F3.4 F3.4-4     F3.4-3 (8/99) - F3.4-4 (8/99)
F3.4 F3.4-6     F3.4-5 (8/99) - F3.4-6 (8/99)
Not Applicable      F3.4-6a (8/99) - F3.4-6b (8/99)
F3.5 F3.5-2     F3.5-1 (Rev. 0) - F3.5-2 (8/99)
    )
F3.6 F3.6-2     F3.6-1 (8/99) F3.6-2 (8/99)
F3.6 F3.6-4     F3.6-3 (8/99) - Blank                                                 )
l Not Applicable      F3 6-3a (8/99) - F3 6-4 (Rev. 0)
F4.1 F4.1-2     F4.1-1 (8/99) - F4.1-2 (8/99)
F4.1 F4.1-4     F4.1-3 (6/97) - F4.1-4 (8/99)
F4.2 F4.2-2     F4.2-1 (6/97) - F4.2-2 (8/99)
F4.2 F4.2-4     F4.2-3 (8/99)- F4.2-4 (Rev. 0) -
F4.2 F4.2-16   F4.2-15 (Rev. 5) - F4.2-16 (2/99)
F4.2 F4.2-18   F4.2-17 (2/99)- F4.2-18 (Rev. 5)
F4.2 F4.2-28   F4.2-27 (8/99) - F4.2-28 (8/99)
F4.2 F4.2-30   F4.2-29 (Rev. 0) - F4.2-30 (2/99)
F4.2 F4.2-32   F4.2-31 (2/99) - F4.2-32 (10/94)
F4.2 F4.2-42   F4.2-41 (Rev. 0) - F4.2-42 (8/99)                                      l F4.2-42a - F4.2-42b F4.2-42a (8/99) - F4.2-42b (8/99)
F4.2 F4.2-44   F4.2-43 (8/99) - F4.2-44 (2/96)
F4.2 F4.2-60   F4.2-59 (Rev. 5) - F4.2-60 (2/99)
F4.2 F4.2-62   F4.2-61 (2/99)- F4 2-62 (Rev. 5)
F4.2 F4.2-72   F4.2-71 (8/99) - F4.2-72 (12/95)
F4.2 F4.2-80   F4.2-79 (4/95) - F4.2-80 (2/99)
,~s     F4.2 F4.2-82   F4.2-81 (4/95) - F4.2-82 (2/99)
F4.2 F4.2-84   F4.2-83 (2/99) - F4.2-84 (2/99)

s Attachment 3    ,
la U-603276       l Page 28 of 29     4 F4 2 F4.2-86 F4 2-85 (Rev. 0) - F4.2-86 (2/99)          .
F4.2-87 04.2-88                F4.2-87 (2/99) - F4.2-88 (Rev. 0)              {
F4.2-9 I - F4.2-92             F4.2-91 (2/99) - F4.2-92 (8/99)
F4.2 F4.2-94              F4.2-93 (8/99) - F4.2-94 (8/99)
F4.2 F4.2-98         _    F4 2-97 (2/95)- F4.2-98 (8/99)
F4.2 F4.2-100  _          F4.2-99 (8/99) - F4.2-100 (6/97)
F4.2-107 - F4.2-108            F4.2-107 (Rev. 0) - F4.2-108 (2/99)
F4.2-109 - F4.2-110           F4.2-109 (2/99) - F4.2-110 (l/95)
F4.2-111 - F4.2-112           F4 2-111 (8/99)- F4.2-112 (8/99)
F4 2-113 - F4.2-114            F4.2-113 (8/99) - F4.2-114 (8/99)
F4.2-115 - F4.2-116           F4.2-115 (8/99) - F4.2-116 (8/99)
F4.2-117 - F4.2-118            F4.2-117 (8/99) - F4.2-118 (8/99)
F4.2-119 - F4.2-120            F4.2-119 (8/99) - F4.2-120 (8/99)
F4.2-121 - F4.2-122            F4.2-121 (8/99) - F4.2-122 (8/99)
F4.2-123 - F4.2-124            F4.2-123 (8/99) - F4.2-124 (Rev. 0)
F4.2-125 - F4.2-126            F4 2-125 (8/99) - F4.2-126 (8/99)
F4.2-127 - F4.2-128            F4.2-127 (8/99) - F4.2-128 (6/97)
F4.2-129 - F4.2-130            F4.2-129 (8/99) - F4.2-130 (5/97)
F4.2-131 - F4.2-132      _
F4.2-131 (6/97) - F4.2-132 (8/99)
F4.2-135 - F4 2- 136 F4.2-135 (8/99) - F4.2-136 (8/99) p.2-139 - F4.2 140             F4.2-139 (8/99) - F4.2-140 (8/99)
F4.2-141 - F4.2-142           F4.2-141 (8/99) - F4.2-142 (8/99)
F4.2-143 - F4.2-144           F4 2-143 (8/99)- F4 2-144 (6/97)
F4.2-145 - F4.2-146           F4.2-145 (9/95) - F4.2-146 (8/99)
F4.2-147 - F4.2-148           F4.2-147 (8/99) - F4.2-148 (4/95)
F4.2-151 - F4.2-152           F4 2-151 (8/99) - F4.2-152 (8/99)
F4.2-153 - F4.2-154           F4.2-153 (8/99) - F4.2-154 (8/99)
F4.2-155 - F4.2-156           F4.2-155 (8/99) - F4.2-156 (6/97)
F4.2-159 - F4.2-160            F4.2-159 (8/99) - F4.2-160 (4/96)
F4.2-161 - F4.2-162            F4 2-16I (8/99) - F4.2-162 (6/97)
F4 2-167 - F4.2-168 F4.2-167 (8/99) - F4 2-168 (Rev. 0)
F4.2-173 - F4.2-174           F4.'l-173 (8/99) - F4.2-174 (Rev. 0)
F4.2-175 - F4.2-176           F4.2-175 (6/97) - F4.2-176 (8/99)
F4.2-l81 - F4.2-182 F4 2-181 (8/99}- F4.2-182 (8/99)
F4.2-185 - F4.2-186                                                             4 F4.2-185 (Rev. 0) - F4.2-186 (8/99)
F4.2-187 - F4.2-188           F4.2-187 (8/99) - F4.2-188 (8/99)
F4.2-193 - F4.2-194 F4.2-193 (11/97) - F4.2-194 (6/97)
FA-I through FA-33            Deletion Page (8/99)
Fig. 4.2.,3.1.10-3            Fig. 4.2.3.1.10-3 (2/99)
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