U-602836, Forwards Rev 7 to Clinton Power Station Usar.Attached Clinton Power Station 10CFR50.59 Rept Identifies All Safety Evaluations for Implemented Changes Affecting USAR for Period of 950430-970630

Forwards Rev 7 to Clinton Power Station Usar.Attached Clinton Power Station 10CFR50.59 Rept Identifies All Safety Evaluations for Implemented Changes Affecting USAR for Period of 950430-970630

ML20217K106
Person / Time
Site:	Clinton
Issue date:	10/20/1997
From:	Jackie Cook ILLINOIS POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20217K110	List: U-602836, Forwards Rev 7 to Clinton Power Station Usar.Attached Clinton Power Station 10CFR50.59 Rept Identifies All Safety Evaluations for Implemented Changes Affecting USAR for Period of 950430-970630
References
GL-96-06, GL-96-6, IEB-96-002, IEB-96-2, U-602836, NUDOCS 9710230230
Download: ML20217K106 (145)

Text

. .

• ,' liknas Power Company

• p e Clinton Power Station P O Box 678 Chnton. IL 61i27 Tel 217 935-5623 Faa 217 9354 532 An litema Company R John o. Oook Senior Vice President U-602836 1 A.120 October 20, 1997 Docket No. 50-461 10CFR50.71(e)

Document Control Desk Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Clinton Power Station Submittal of the Updated Safety Analysis Report. Revision 7

Dear Madam or Sir:

Pursuant to 10CFR50.71(e), Illinois Power (IP) hereby submits one original and ten copies of Revision 7 to the Clinton Power Station (CPS) Updated Safety Analysis Report (USAR).

In accordance with 10CFR50.71(e)(2)(ii), the attached Clinton Power Station 10CFR50.59 Report (Attachmer.t 2) identifies all safety evaluations for implemented l changes affecting the USAR for the period of April 30,1995 through June 30,1997.

Tlase safety evalunions were performed by IP under the provisions of 10CFR50.59.

a .

The revised USAR pages (Attachment 3) are annotated with Revision 7 and revision bars. Subsequent revisions of the USAR will be submitted within six months of the completion of a CPS refueling outage or within 24 months from the date of this submittal and will reflect a'l changes up to a maximum of six months prior to the date of filing.

/

, fos S y Jssc: '

i 9710230230 971020 l

r" a " o'

= spWE8EM l

6=. ' l . ,. U-602836

- Page 2 As required by 10CFR50.71(c)(2)(i), the information given in the attached USAR

- accurately presents changes made since the previcus submittal and analyses submitted to the Commission or prepared pursuant to Commission requirement. Notwithstanding, IP acknowledges that CPS USAR discrepancies do exist. Known USAR discrepancies are evaluated, tracked, and corrected via the CPS conective action program.

Sincerely yours, f

f J.G. Cook Senior Vice President LBM/krk Attachments cc: NRC Clinton Licensing Project Manager NRC Resident Office, V-690 Regional Administrator, Region III, USNRC Illinois Department ofNuclear Safety

= . .

Attachment I to U 602836 J. G. Cook, being first duly sworn, deposes and says: That he is Senior Vice President of the Nuclear Program at Illinois Power; that the Clinton Power Station Updated Safety Analysis Report, Revision 7, has been prepared under his supervision and direction; that he knows the contents thereof; and that to the best of his knowledge and belief said letter and the facts contained therein are true and correct.

Date: This 1O day of October 1997.

J.G. Cook

:::::: z :::

STATE OF ILLINOIS l SS. *NE' Junemene8.MasNas

/- b , NotaryMas,amesof tanels ul4+* COUNTY J  ; MyCommission E$res11/2W ,

Subscribed and sworn to before me this2O day of October 1997.

c'_0. .n,6 c5 YLLO

/g (Notary Public)'

Attachment 2 to U-602836 Page 1 of142 CLINTON POWER STATION 10CFR50.59 REPORT

Attachment 2 to U-602M6 Page 2 of 142 -

i ADDITION OF GAUGES TO MONITOR FRESSURE AND CHANGE IMC018 TO

NORMALLY CLOSED.

Document Evaluated: ECN 28660 Log Number: 89-0082 This change was previously reported for USAR Revision 2 as Temporary Modification (TM)

,89-033. The Engineering Change Notice (ECN) changed the temporary configuration to permanent, The modification revised the normal position of the Makeup Condensate (MC)

system valve from open to closed. This valve is used to provide makeup water to the Standby Liquid Control (SC) System. This action was taken because MC water had been leaking past the L SC tank isolation valves which would dilute the borated water out of specification and raise tank level above the allowable level. A pressure gauge was also mstalled on a low point drain in the SC system to provide for periodically monitoring the pressure in the SC system piping to verify that the SC pumps were adequately primed with the MC valve closed. The pressure gauge did ,

not significantly affect the integrity of the SC system piping as the drain valves located between i the gauge and the piping were only opened while the pressure was being checked.- Sampling of

'the water in the SC system piping was also performed to ensure that sodium pentaborate solution

in the SC tank did not leak past the tank's isolation valves without being detected, i

i~ RETIREMENT OF CONTROL ROOM CHLORINE DETECTORS

{ ' Document Evaluated: Modification VCF013 Log Number: 91-033

- This change retired in place the Control Room Ventilation (VC) System chlorine detectors and removed the "High Chlorine" annunciators in the main control room.' The purpose of the chlorine p detectors is to provide an alarm in the control room, upon detection of high chlorine concentration in the VC makeup air, to alert operators to place the VC system in the chlorine i mode of operation to prevent entrance of chl >rine into the control roomi The requirements of

Regulatory Guides 1.78, " Assumptions for E valuating the Habitability of a Nuclear Power Plant

[ Control Room During a Postulated Hazardous Chemical Release," and 1.95, " Protection of

{ Nuclear Power Plant Control Room Operato rs Against an Accidental Chlorine Release," do not ._

! . apply for three specific reasons. Clinton Power Station (CPS) limits the size of chlorine j; containers on site to leu than 100 pounds. No other chlorine storage sites are located within five

! miles of the plant. Transportation surveys sltow that the frequencies of chlorine shipments within L five miles of the plant are below the values a llowed by Regulatory Guide 1.78. These surveys are

_ perfonned every three years. In addition,11 e CPS Technical Specifications (TS) have already been amended (Amendment 12 granted Nos ember 12,1988) to delete chlorine monitor

, - requirements as long as no chlorine containxs larger than 100 pounds are located on site. As

such, this change has been adequately reviewed without an increase in consequences or new i- accident types anticipated.

F h

. + - , . , - ,e,w,., , - - - --- ,, - - - ,- ,,,- , ,--+-w - , - - , ,

v

- . , _ ~ . . - - - . - - - . - - - - - - - ,

- Attachment 2 to U402836 - l Page 3 of142 1 1 .

l SELF-LUBRICATING BUSHINGS FW DIVIQON III SX PUMP Document Evaluated: Modification SXF022 - LogNumber: 92-097 j

h mt modification SXF022 allowed use of self-lubricating bushings and Belzona materials in the S.iutdown Service Water (SX) System Division III pump. Previous cutless rubber bushings required water, supplied from the lake, as a lubricating medium. Impurities in the lake water L

caused silt to build up in the bushings causing pump binding. The use of self-lubricating bushings eliminated this problem and enhance pump dependability. These self-lubricating bushings are already in use in the Division I and II pumps without having had adverse effects. Replacement of existing bushings eliminates the requirements of associated lubricating piping, valves and -

supports.- This will remove the lubrication piping shown on USAR Figure 9.2-2, Sheet 3 and

- - Figure 3.61, Sheet 49. There are no failures regarding pump bushings or the lubricating system L

discussed in the USAR. This has been evaluated for seismic and environmental qualification

!- ' concerns with no adverse impact determined. No new failures are created by the change since the

i. equipment piping has been evaluated for seismic and environmental impacts and two other SX p system divisions already use this material.

4 REA' CTOR RECIRCULATION (RR) PUMP VIBRATION MONITORING i

} Document Evaluated: Modification RRF015 . Log Number: 93-045 R1 This modification makes permanent two temporary modifications (TMs) previously installed. TM

{

031 reported for USAR Revision 3 installed a vibration monitoring system on the reactor

[ recirculation 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 lifting lugs using welded support blocks. The accelerometers measure pump casing vibration.

The velocity transducers are mounted on support blocks welded to the top of the motor casing

. and measure lateral motor casing vibration. TM 92-007 later installed two horizontal proximity probes and axial proximity probes on each pump. These are also to monitor pump vibration.

4 RRF015 changes these two TMs to permanent status and installs a personal computer in the l

?- control room computer room to provide continuous monitoring and analysis of vibration data for i the RR pumps and motors. This modification changes USAR Appendix E and F fire protection

. information. The installation was evaluated for seismic loads, divisional separation, and fire F loading and was found acceptable. This modification provides a monitoring function only; no trips or interlock permissives are associated with the operation of the vibration monitors. No new 1

- failures or accident types are created by the change since the new equipment is passive in nature i'

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 7. However, portions have been installed.

[

- -An updated status of this change will be reported in Revision 8 to the USAR.

I r

i i

a # s' fY I e - ,

~ . - _ - - - - - . . . _ - . - - . - .- - . - - - - .- - - .

Attachment 2

- to U-602836 Page 4 of142 -

- -ACTUATOR AND SUPPORTING HARDWARE REMOVAL FROM ECCS INJECTION j- CHECK VALVES Document Evaluated: Modification M-079 LogNumber: 93-114-

- This modification removes the actuator hardware from the Emergency Core Cooling (ECCS)

. . System testable check valves. This change prevents the remote operatiom ' valves IE12F041 A,-

1 lE12F041B, IE12F041C,1E21F006, and IE22F005 during power operm This remote operation capability was to satisfy a portion of the ASME Code,Section IX. An ASME Code,

: Section XI, relief request (2014) was submitted and approved by the NRC on September 13; 4

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; the probability of failure I

or likelihood of accidents involving these valves would not increase. This change has not been fully implemented as of the cutoff date for US AR Revision 7. However, portions have been -

. installed. An updated status of this change will be reported in Revision 8 to the USAR. USAR L text and figures are being revised as portions of this modification are released for operations -

Reference to check valves being " testable" will not be changed as they can still be locally tested.

Reference to the valves being " air-operated" will be changed since actuators are being removed.

ELIMINATION OF THE STEAM CONDENSING MODE OF RHR OPERATION Document Evaluated:. Modification RH 033 LogNumber: 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

- restarted. This would be done by directing main steam line steam through a Reactor Core Isolation Cooling (RI) System line into one of the two RH heat exchangers. The startup testing g _ of this mode of operation was deleted by safety evaluation 87-1353 and reported in USAR Revision 0. As such, removing this as a startup test requirement and as a mode of operation at (1 CPS did not decrease the margin of safety or increase the consequences of an accident. In .

addition, Temporary Modification GO-04 was reported for USAR Revision 2. This temporary _

L modification removed the piping spool between the RI syr m piping and the RH heat exchangers.

l . RH-033 makes this change permanent. This change will a; low seven pipe hangers to be ' changed from snubbers to struts. -The change provides for the determination of twelve Motor Operated Valves (MOVs) at the Motor Control Center (MCC) and de-terminating of cables entering the Main _ Control Room (MCR). RH-033 includes extensive USAR changes to eliminate 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 accident.

This change has not been fully complete as of the cutoff date for USAR Revision 7. However,

! portions have been installed. An updated status of this change will be reported in Revision 8 to Qe USAR. USAR text and figures are being revised as portions of this modification are released for operations.

l l

- _.._ , __ _ . , _ .._ . _ ._ ._ _ _ _~

t Attachment 2 to U 602836 .

Page 5 of142 e

MODIFICATION TO NSPS INVERTERS Document Evaluated

Modification IP-F004 LogNumber: 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 DU system voltage _

transients by increasing the capacitance of the input filter 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 change improves 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.

i 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 7.4 An updated status of this change will be reported in Revision 8 to the USAR.

EXTEND NITROGEN CONNECTION OUTSIDE HEPA FILTER ROOM i.

Document Evaluated: Modification OG-F016 Log Number: 94-036 l

Modification OG-F016 extends the nitrogen purge piping for the Offgas Treatment (OG) System HEPA filter room from inside the room to an existing connection outside the room. The modification is being installed to increase accessibility to the nitrogen purge connection in the E

event of an OG charcoal adsorber fire. This change also reduces radiation exposure to plant

personnel if a nitrogen purge is needed. The change will not increase the probability of an offgas system failure.

This change was reported in Revision 6 to the USAR. At that time, the change had yet to be started. However, since then, the change has been fully implemented.

INSTALL SMOKE DETECTOR IN DIVISION 3 BATTERY ROOM.

Document Evaluated: Modification FP-089 - Log Number: 94-050 Modification FP-089 resulted in the installation of an ionization smoke detector in the Division 3 F . battery room as corrective action for Condition Report 3-91-10-040. During the review offire i load calculations, it was found that non-conservative assumptions were made in determining the f quantities of combustible materials. To comply with 10CFR50, Appendix R and the CPS safe shutdown analysis, an autom'atic fire detection system is necessary because of the increased fire

( . loading. The detector is installed in a non_-safety related circuit; therefore, no new loads are j- placed on a safety related circuits. The detector and associated electrical conduits are seismically j mounted to prevent any damage to safety related equipment. No new failures have been created i by this change.

J r e - - - - - - - - - m- - - - - -

4 Attachment 2 to U-602836 Page 6 of142

[ This change was reponed in Revision 6 to the USAR. The change was not fully implemented within the time frame specified in the USAR Revision 6 letter. However, since then, the change

, has been fully implemented.

i PRESSURE LOCKING FREVENTION FOR LOW PRESSURE ECCS INJECTION VALVES Document Evaluated: Modification M-082 Log Number: 94-072

! Modification M-082 modifies low pressure Emergency Core Cooling System (ECCS) injection _

~ valves, IE12F042A, IE12F042B, IE12F042C, and IE21F005, by drilling a hole in the valve disk.

.' This modificath,a prevents pressure locking and thermal binding as identified in General Electric t (GE) Service Information Letter 368, " Recirculation System Isolation Valve Locking," NRC

. Circular 77-05, " Liquid Entrapment in Valve Bonnets," and NRC Information Notic ' l-31, t

" Failure of Safety Injection Valves to Operate Against Differential Pressure," and h-26,

" Pressure Locking ofMotor-Operated Flexible Wedge Gate Valves." Valves IE12F042A, l IE12F042C and IE12F005 were modified in the fifth refueling outage, RF-5. IE12F042B was L completed in RF-6. No new failures have been created by the change. The change eliminates a potential failure of disabling one or more of the injection valves.

This change was reponed in Revision 6 to the USAR. At that time, the change was not fully I

implemented. However, since then, the change has been fully implemented.

I CONDENSATE FILTRATION SYSTEM MODIFICATION l Document Evaluated: Modification CP-020 Log Number: 94-075 R1 Modification CP-020 changes the USAR description of the Condensate Cleanup System (CCS) by introducing the Condensate Filtration (CF) System as a subsystem. The CF system filters remove

non-soluble impurities, mainly insoluble iron, from the condensate. This is intended to reduce 4

occupational radiation exposure and reduce low level radioactive waste volume. This will be done by placing a condensate filter vessel upstream of the A, B and C deep bed demineralizers.

[ . Each fiher and demineralizer are enclosed in'a dedicated cubicle. All three cubicles will be

4.  : installed and released under partial releases. Isolation of cubicle A of CP-020 will have no l= adverse impact on plant operation because the CCS can be isolated without removing interfacing

' systems from operation. The same is true for future portions of this modification.

I CCS piping and vessels, including the CF system are located in the non-safety-related Turbine and Radwaste Buildings. Liquid line breaks outside containment are bounded by the feedwater line L break. No new failure modes have been introduced by this change. The accident analysis in e USAR Chapter 15 was reviewed and it was determined that this change does not introduce operation outside the bounds of any analyzed condition.

This change has not been fully complete as of the cutoff date for 'USAR Revision 7. However,

[ portions have been installed. An updated status of this change will be reported in Revision 8 to r the USAR.

I; i

i, -_ , . . _ , _ _ _

,_m.. _ _ . . _. . , _, . . ._ , ..__ _ -

Attachment 2 to U402s36 Page 7 of 142 ;

INSTALL PERMANENT GAGGING DEVICES ON STEAM JET AIR EJECTOR (SJAE)

CONTROL VALVES Document Evaluated: ECNs 28912,28913 ~ Log Number: 94-086 This change installs a permanent mechanical gagging device;on steam jet air ejector (SJAE) control valves, IB21F437A and IB21F437B, to close the valves at all times during SJAE-operations. These valves are designed to be used during SJAE startup at low operating pressures

'(less than 185 psig). Since the SJAE's are normally put into operation at pressures greater than 185 psig, the valves do not perform any useful function. Mechanically locking the valves closed

> will prevent the valves from inadvertently opening following a loss of air or actuator failure, thus ,

preventing an unplanned shutdown. No new failure modes are created by this change.

This change was reported in Revision 6 to the USAR The change was not fully implemented ~

within the time frame specified in the USAR Revision 6 letter. However, since then, the change has been fully implemented.

REGULATORY GUIDE (RG) 1.58, " QUALIFICATION OF NUCLEAR POWER PLANT INSPECTION, EXAMINATION, AND TESTING PERSONNEL" Document Evaluated: USAR Section 1.8, RG 1.58 Log Number: 95-025

CPS's compliance with Regulatory Guides is delineated in USAR Section 1.8. This USAR change

- clarifies how Level II and Level III personnel demonstrate cenain capabilities as described in ANSI N45.2.6-1978, " Qualifications ofInspection, Examination, and Testing Personnel for Nuclear Power Plants,"_ which is referenced in RG 1.58, " Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel." The USAR is being revised to state that Level 11 and III personnel will be required to demonstrate functional capabilities only to the extent that those functions are performed. The functional capabilities will be documented on inspector certification records. :This clarification to inspector personnel has no impact on failure probability.

i No credit is taken for either Level II or III inspector qualifications in USAR failure analyses.

USAR APPENDIX E AND F CHANGES NOT IDENTIFIED FOR PREVIOUS DESIGN CHANGES Document Evaluated: CR 1-93-06-027 Lng Number: 95-040 This Condition Report (CR) was issued to correct USAR text and figure changes which had not been done for plant changes made from 1986 to mid-1993. These changes were done prior to commercial operation of CPS or had already been evaluated and reported in USAR Revisions 0 through 5 with one exception. As such, except for the change described below, these changes have already been evaluated and a determination made that no unreviewed safety questions were involved.

In 1991, two restrooms in the Turbine Building at elevation 721' (fire zone T-1 A) and elevation 762' (fire zone T-lh) were removed. The rooms were part of the non-safety related Turbine

[ -

Attachment 2 to U-602836 Page 8 of142

^

. Builuing. . In removing the Turbine Building restrooms, the sanitary drains were plugged to avoid a possible non-rnonitored release path. - Removal of bathroom fixtures and other internals did not -

affect the integrity of the rooms, which are now used for general storage. During removal of these restrooms, the USAR figures in Appendix E and F that showed these rooms were not identified. The USAR changes were made to update the drawings in Appendices E and F to reflect the current use of these rooms.

CHANGE WS SYSTEM RELIEF VALVE SETPOINTS Document Evaluated: ECNs' 28984, 29000, 29001, 29002 -Log Number: 95-042 This change was reported in Revision 6 to the USAR. At that time, the changes had been approved but had not begun. Since then, the change has been fully implemented. The change revises the setpoint for Service Water (WS) System relief valves, lWS166A,1WS166B, IWS153A, and IWS153B, to be consistent with the maximum allowable working pressure for the

turbine oil cooler and generator stator cooling water heat exchangers. The relief valves will still

provide thermal relief protection for the heat exchangers. The changes are corrective actions for  ;

l condition report 1-94-09-062. No new failures or accident types are expected since the system l - will function as designed.

[ CONTAINMENT WEIR BOX FLOW INDICATION REMOVAL i

Document Evaluated. ECN 29135 LogNumber: 95-053-i

[ ' This change was reported in Revision 6 to the USAR. At that time, the change was partially F completed. Since then, all portions of the change have been fully implemented, - This change removes from operation the containment floor and equipment drain weir box flow measuring

. devices and takes credit for alternate methods ofleak detection. The weir box flow measuring

[ devices (v-notch plates) are located in containment and are high maintenance devices due to clogging. The clogging has caused high flow indications and false alarms. Alternate methods to p

!  : determine leakage are monitoring containment equipment and floor drain sump fill and run timers e and flow totalizers. The alternate methods will assure compliance with Regulatory Guide 1.45, l

" Reactor Coolant Pressure Boundary Leakage Detection Syste,ns."

CHANGE CP SYSTEM RELIEF VALVE SET POINTS Document Evaluated: ECNs 29126,29127 and 29128 LogNumber: 95-054 The change revises the setpoint for Condensate Polishing (CP) System relief valves,1CP074, ICP075, and ICP013, to be consistent with the maximum allowable working pressure of tanks OCP0lT, OCP02T, and ICP06T respectively. The previous setting was 105 psi. The new setting

is 100 psi. The changes will not impact how the system or components in the CP system operate j because thermal reliefis a factor only during abnormally high room temperature environmental

, conditions. The changes are corrective actions for condition report 1-94-09-062. No new failures or accident types are expected since the system will function as designed. In addition, this

l Attachment 2 to U.502836

[ _ _ Page 9 of142 -

- will better protect these tanks as the new settings will prevent a potential condition were the e operating pressure would be up to 5 psi over the design pressure.

l-

. ANSI N509/510 VARIANCE REPORT, USAR SECTION 1.8 l

l Document Evaluated: ANSI N509/510 = LogNumber: 95-055 This change included the ANSI N509/510 Variance Report as a referenced clarifisation and

[

exception document for Regulatory Guides 1.52, Revision 2, " Design, Testing, and Maintenance Criteria for Engineered Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption

- Units of Light-Water-Cooled Nuclear Power Plants," and 1.140, Revision 0, " Design, Testing, and Maintenance Criteria for Normal Ventilatien Exhaust System Air Filtration and Adsorption

~~

Units of Light-Water-Cooled Nuclear Power Plants," as defined in USAR Section 1.8. ANSI N509/510 defines specific test acceptance criteria. This report was provided to the NRC for

review as part ofNRC open item 461/85004-08. Upon review, the NRC found the " Minor j variances appear to have adequate engineeringjustification." Each variance is unique and justified -

as acceptable in that individually or collectively, these " minor variances" to N509/510 will not p impact equipment design basis function; I ELIMINATION OF DG OPERABILITY FOR ECCS OPERABILITY DURING A PLANT SHUTDOWN

Document INaluated

TS 3/4.5.2 USAR Change LogNumber: 95-058 i'

[

This change is to eliminate the requirements previously contained in Technical Specification (TS) 3/4.5.2 that an Emergency Core Cooling System (ECCS) subsystem be capable of being powered -

- from an OPERATIONAL Diesel Generator (DG) for the ECCS subsystem to be considered

! OPERABLE during MODES 4 and 5. This requirement was not retained in the CPS Improved L Technical Specifications in Amendment 95. The NRC Safety Evaluation (SE) fo* TS Amendment

- 95 states, "Unless otherwise specified in this safety evaluation, the relocated limiting conditions L for operation (LCO) portion of the existing TS, which includes the system description, design

limits, functional capabilities, and performance levels, will be relocated to the USAR." Rather

than relocating this requirement to the USAR, Illinois Power (IP) decided to delete these

[ - requirements by evaluating this as a USAR change.

l The TS Bases function of ECCS in MODES 4 and 5 is to ensure adequate coolant inventory and

- sufficient heat removal capability for the irradiated fuel in the core in the event of an inadvertent i reactor drain down. An inadvertent reactor drain down is not assumed or analyzed in the USAR.

Therefore, whether a DG is capable of supplying power to the ECCS subsystem does not alter any malfunctions of the associated ECCS subsystems previously evaluated in the USAR. USAR

Chapter 15 accidents were reviewed and only USAR_Section 15.7.4 was applicable to shutdown 4

conditions. However, this section did not require ECCS systems for mitigation. Therefore, the

- probability and consequences of an accident were not changed. In addition, TS LCO 3.4.10 and 3.9.9 requires two Residual Heat Removal (RH) shutdown cooling subsystems to be OPERABLE in Gese MODES, one of which must, by default, be capable of being powered by an OPERABLE DG. The SE for Amendment 95 also allowed a second offsite circuit to be available instead of the 4

m ,,-e -.-+ -. a ., - - - - - , , - , , - , , . - - ,

Attachment 2 '

to U 602836 Page 10 of142 -

requirement for the Division III DG to be OPERABLE when the High Pressure Core Spray - .

3 (HPCS) System is required to be OPERABLE. This was allowed based on offsite circuits being at least as reliable as a DG. ,

CORRECT ERROR FOR POINT OF CHEMICAL INJECTION ON ECN 28885

Document Evaluated

ECN 29155 Log Number: 95-059.

This ECN augments the safety evaluation (SE) for SE 94-0080, reported in USAR Revision 6.

! SE 94-0080 converted Temporary Modification 93-017 to a permanent modification which provides an injection system to add a scale-inhibiting chemical to the Circulating Water (CW)

' System.- The purpose of the chemical treatment is to prevent further scale build up on the main condenser tubes. The chemical feed equipment is located in the circulating water pump house and

< . injects chemicals into the discharge flow of the "A" or "C" CW pump. The chemical treatment

~

F does not require changes to the National Pollution Discharge Elimination System permit or l approval by the Environmental Protection Agency. ECN 29155 corrects the chemical injection points to the as shown on USAR Figure 10.4-3 No new failures were determined to have been

, introduced by the change and the additional fire load was evaluated to not affect the safe

shutdown analysis. This change in location also did not impact the chemical injection rates, CW

flow rate or the injection tubing itself.

. USAR FIRE PROTECTION SUBSECTION REVISION 1

Document Evaluated: USAR Section 9.5.1.1 LogNumber: 95-061 '

l- This' change adds a clarification to USAR Section 9.5.1.1, " Design Bases," of Section 9.5.1, " Fire Protection System." The clarification states that the CPS document "NFPA Code Corformance i Evaluation With Completed Checidists," was prepared prior to CPS entering into commercial operation, identifies the level of compliance of the CPS fire protection program with National Fire Pr otection Association (NFPA) codes and, where appropriate, provides technicaljustification for deviations based on the fire protection equipment and procedures which existed at that time. This

i. . document is only kept current regarding documentation of non-conformance or partial conformance with NFPA codes. This document was needed prior to the NRC's Appendix R fire l

protection inspection of CPS. This was performed prior to the issuance of the original CPS Operating License. Other than keeping this document current for current code deviations, it is i

used as a reference document. Therefore, this USAR clarification was necessary. This clarification will not cause accidents or create conditions not analyzed. CPS fire protection

, . changes will still be evaluated for unreviewed safety questions.

ABANDON IN PLACE SELECT DRYWELL FILTER TRAIN "C" DRAIN LINES 9

Document Evaluated: Modification VQF006 Log Number: 95-062 The inlet and outlet drains for the moisture separator, prefilter and HEPA filter drywell purge filter, Train "C" of the Drywell Ventilation (VQ) System are clogged with concrete. This modification abandons in place these drain lines and changes USAR Figure 9.4-13, Sheet I to y

Anachment 2 to U402s%

Page 11 of142 reflect this change. This does not affect how the equipment operates. The only entrained water enters upstream of the moisture separator which is designed to remove entrained moisture. If this would occur, as in the case of a large steam line break, the "A" or "B" independent trains could be put into service.L Therefore, this would not impact equipment malfunctions. This does not reduce margin of safety as defined in the basis of any CPS Technical Specification in that tne

inability to use these drains does not detrimentally affect the operation of the drywell purge filter

train. -

REVISION OF SPN SETPOINT DETERMINATION (ACOUSTIC MONITORING)

Document Evaluated: USAR Section 7.6.1.11 LogNumber: 95-063 The USAR describes two methods of Safety Relief Valve (SRV) acoustic monitor setpoint

- determination methodology. This change revises this section to change these methods to alternatives. The change also reflects that when using the calculation method and spurious alarms occur on background noise, the gain may be reduced to eliminate spurious alarms. This change is in accordance with vendor recommendations for methods ofinitial gain point settings. Since this change still has this equipment operating in accordance with original vendor design, there is no -

change in probability of equipment malfunction. There are no design bases accidents specifically requiring acoustic monitoring to be operable. There has not been a change in accidents or accident types since there has not been equipment utilization changes.

POST ACCIDENT SAMPLING SYSTEM (PASS) TRAINING CHANGES Document Evaluated: USAR Changes Log Number: 95-064 R1 USAR Sections 7.6.2.12.5.9 and 9.3.7.5 refer to the training program instituted for the Post-Accident Sampling System (PASS). This section previously required an operator to complete se.ni-annual refresher training as well as biennial requalification. Changes have been made in the PASS training program such that refresher training will be used to maintain operator skills without the need for the requalification training. This change removas mention of requalification training, adds clarification of how refresher training can be conducted and removes the redundant reference in Chapter 7. This change is to training conducted for operation of the PASS. It does not change the system or how it is operated. Therefore, the changes would not increase the probability of an accident, create new accidents not already evaluated in the USAR or cause an unreviewed safety question.

OPERATING PROCEDURE FOR LIQUID VOLUME REDUCTION SYSTEM AT CPS Document Evaluated: OM-098-WS Log Number: 95-065 A CPS specific Radwaste Volume Reduction (RVR) process control procedure that incorporates plant specific terminology and program requirements has been established. The Process Control Program is described in USAR Section 11.4.2.4, and the new processing procedure constitutes a change to this section. This procedure addresses the processing of concentrated waste on a case by case basis using the RVR process. This' process is dependent upon waste stream activity 1

_ .m

lI ' Attachment 2 to U402sM Page 12 of142 -

i, concentrations which are cvaluated per the requirements of the site 10 CFR Part 61 sampling '

program.' This RVR system uses a blender and a dryer to remove the free liquid from the solids of 1 the waste stream; When the dry powder product is formed, it is placed in a high integrity e - container (HIC) for shipment and burial. The potential for spills and releases in the Radwaste -

I Building have been previously analyzed. Although this change represents a change to the method

of processing concentrated waste, no additional radioactive source term'or physical form is being

- introduced by this change.- Usage of the Turbine Building crane to move the HICs is in e accordance with USAR Section 11.4.2.6 and has been evaluated and previously reported in p cUSAR Revision 6 in safety evaluation 94-045.

POWER SUPPRESSION TESTING PROCEDURE AND DATA SHEET

[

i

!- Document Evaluatedt CPS Procedore 2503.02 & Log Number: 95-066 i: Data Sheet 2503.02D001 This ii a new procedure and data sheet used to collect data for use in determining the location of leaking fuel assemblies. The procedure directs that reactor power be reduced to 60-65% to minimize further damaging the failed fuel; One rod will be inserted at a time for approximately 15 minutes to permit accurate assersment of the radio-chemical response. The' Rod Withdrawal Limiter (RWL) will then be bypassed on this individual rod. The rod will then be continuously withdrawn to the original location. The off-gas activity changes will be recorded durirg this procedure on a chart recorder temporarily connected to the off-gas pre-treatment monitor.

No failures of the off-gas pre-treatment monitoring system are discussed in the USAR. This

- monitor provides indication onlyi Temporarily installing a chart recorder would not impact the monitor. The procedure requires rod insertion and withdrawal while monitoring the chart records. Rod Withdrawal Error (RWE) and Minimum Critical Power Ratio (MCPR) are not impacted by this operation per Analysis of General Electric (GE) document 22A7007. Prevention of the Linear Heat Generation Rate (LHGR) reaching a point of 1% plastic strain was also reviewed but MCPR was determined to be the most limiting safety criterion. The generic Boiling Water Reactor BWR-6 RWE analysis determined the allowable withdrawal increments. These-withdrawal limits are physically implemented by the RWL function of the Rod Pattern Control -

. System (RPCS). When a rod can be withdrawn past the original position, this procedure requires e specific RWE analysis. This aralysis show that the original BWR-6 RWE analysis is bounding prior to step performance. Additional controls to prevent an RWE include procedural requirements to not withdraw the bypassed rod beyond the original position and requires that a second licensed operator verify proper rod movement; These safeguards ensure that the USAR Section 15.4, " Reactivity and Power Distribution Anomalies," are not negatively affected.

PORTABLE AND LABORATORY TECHNICAL EQUIPMENT AND INSTRUMENTATION Document Evaluated: USAR Table 12.5-2, E-Plan Table 3-2 Log Number: 95-067 The USAR describes the location and type of fixed and portable area and airborne radioactivity monitoring equipment to be used during normal, abnormal and following accident conditions to keep occupational dose ALARA. This change better defines where this equipment can be located.

t Attachment 2 to U.602836 -

l' age 13 of142 In' addition, the evaluation changes the accuracy requirements and instrument locations for the high-range survey meters.- The USAR states the instrument accuracype t 5%. USAR Table 12 5 2 and Emergency Plan (E-Plan) Table 3-2 are being cl.anged to reflect the actual instrument

= ac<uracy ofi 10%. These tables are also changed to add the Radiation Protection Calibration Facility as a location for these instruments. The change to t10% is per the recommendations of

AN SI Standard N320-1978, " Performance Specification for Reactor Emergency Radiological Monitoring Instrumentation." Therefore, this change would not impact equipment malfunctions, y

acciilents nor create the potential to increase the severity of an accident. This is equipment used for CPS personnel ALARA concerns.

REMOVE TEMPERATURE TOLERANCES FROM VL AND VW SYSTEMS Docur.)ent Evaluated:-- ECN 29284 LogNumber: 95-069 This ECN revises the design condition documents and USAR for the Laboratory Ventilation (VL) and Ra6vaste Building Ventilation (VW) Systems. The ECN removes the requirements that the ,

tempera ure in the laboratory be maintained at 75 i 1* F, and the counting room be maintained at ,

70 i l' 12. This also removes the requirements that the temperature in the Radwaste Operation Center (R OC) be 73 i l' F and the storeroom and C&I calibration lab areas be maintained at 75 i P F. These numbers have been changed in the USAR to "approximately 75'F" for the VL -

system ar6"approximately 73'F" for the VW system. The intent of these systems is to maintain the areas ti.cy serve in a comfortable climate for the personnel and equipment in the ann. The removal of the tolerances does not change this intent. Prior to significant environmental changes,

- changes whi:h would impact equipment qualifications, personnel working in the area would respond. Th3 VL and VW systems do not have any equipment malfunctions evaluated in the '

USAR. These are non-safety systems not required to assure either the integrity of the reactor coolant boundary or the capability to shut down the reactor and maintain it in a safe shutdown condition. Therefore, this change would not impact the margin of safety as described in the USAR.

RAW WATERTREATMENT SYSTEM Document Evaluated: Modification CL-007 Log Number: 95-070 This change adds taps in the Shutdown Service Water (SX) System, injection quills in the

. Circulating Water (CW) System sample connections in the seal well for the Service Water (WS) and CW Systems, connects unused piping in the Chlorination (CL) System and utilizes existing diffusers in the pump suctx.n areas of the WS and CW systems. This modification also installs foundations for buildings or tanks and valve pits at the seal well or south of the Control Building.

Additionally, the modification connects non-Class IE electrical power to accomrmdate supplied, built, operated and maintained raw water treatment system components. This treatment system is intended to reduce fouling, corrosion and plant discharge violations. This is done by adding scale inhibitors, sodium hypochlorite and codium bisulfite at various locations.

_._.__..___,.__.__m _ ._._ _ .. _ ..- _.. _ _. _ _ _ _. _.._ _ _ _

)

Attacionent 2 a- '

- to U 602816 0 .. Page 14 of 142 1

( . Equipment installatlon has been' designed and qualified to that of the original equipment design.

i The changes to equipment important to safety are limited to the addition of pipe taps for injecting i- . sodium bisulfite into the SX discharge headers. This change was evaluated for structural integrity

-and for adverse chemical reactions. The change will not increase any accident probability as the 1 CW, WS, SX and Non-lE electrical distribution systems are not initiating events for any accident evaluated in the USAR.

l

[- - WX TANK LEVEL MODIFICATION - SOLID RADWASTE SLUDGE TANK (0WX02TA) i Docutaent Evaluated: Modification WX-027 Log Number: 95-072

= .

1 Modification WX-027 replaces the ultrasonic continuous level indicating system on the Solid <

Radwaste Processing (WX) tank OWX02TA with on-demand plumb bob level measuring

. Instrumentation to measure buh total tank level and sludge (interface) level. The modification p dso installs a " Thermal Dispersion" RTD type sensor element and circuitry for a continuous high-l high level alarm. In addition, the pump trip and high and low sludge level alarms are being s removed due to the manual nature of the new level monitoring system. The pump cannot be running due to the likelihood ofit damaging the plumb bob. This modification is similar to WX-

~ F010, reported in Revision 6 to the USAR. USAR Figure 11.4-3, Sheet 5, Appendix E Figures

]. FP-16a & b, Appendix E Cable Tray Drawing sections 14,7.7.1.12.3.1 and 1.8 for Regulatory F Guide 1.143, " Design Guidance for Radioactive Waste Management Systenu, Structures, and

. Components Installed in Light-Water-Cooled Nuclear Power Plants," are being changed by this

. modification. This modification is being made to improve the reliability of the level monitoring i instrumentation for this tank. - The only associated safety related failure evalanted in the USAR is -

j WX tank rupture. This change does not impact the probability or consequences of the WX l_ rupture event.

i:

p SX HEAT EXCHANGER PERFORMANCE TEST PROCEDURE CHANGES I .

l Document Evaluated: CPS Data Sheet 3602.01D007 LogNumber: 95-073 This evaluation is for CPS Procedure 2602.01, " Heat Exchanger Performance of Shutdown

Service Water Coolers Covered by NRC Generic Letter 89-13," data sheet 2602.01D007, " Heat Exchanger Flush," This data sheet is being created to provide a location to record applicable information such as throttle valve identification number / position, maximum flow achieved, as 4

found/as left flows, etc. The data sheet is used when flushing safety related heat exchangers

supplied by the Service Water (WS) and Shutdown Service Water (SX) Systems. Heat Exchanger (HX) flushing is being done to reduce the amount of mud / silt buildup in the applicable h heat exchangers. This will help ensure that the HXs are capable of performing their intended

design function.- This flushing is accomplished by allowing the maximum amount of water / highest

[ flowrate to be flushed through the HXs for approximately two hours. This procedure change and

[ ' flushing activity does not change the function or design of the affected equipment / systems. The

[  ; main impact is decreased flow to other HXs when full flushing flow through a HX is established.

h  : This has been evaluated and due to the low flows (15 to 125 gpm) of the HXs affected by this, j_ other applicable HXs will be able to perform their intended functions. Due to the short length of

(

L time the flushing is conducted, erosion / corrosion degradation is expected to be insignificant.

i 1

i,. . - . _ _ -_ _ . , , _ _ _ - - - - - - ___ _. _ . . _ ., _ _ ~ _.__ - , -

l J Attachment 2 -

to U 602P36

' Page 15 of142 REMOVAL OF THE DIVISION III STATOR TEMPERATURE INDICATOR-Document Evaluated: ECN 28851 - Log Number: 95-074 -

This ECN allows the removal ofinstrument ITI-DG253 which is the generator stator temperature indicator for the Division III Diesel Generator (DG). The instrument provides temperature ir.dication for the stator as well as provides annunciation. Nwever, it does not provide for any i trips or interlocks to the DG, This change is being made because this indicator is no longer available. This change affects the USAR by requiring a change to Section 8.3.1.1.2. This section lists alarms that are at the Division III local panes. No probability or consequences of a malfunction are affected as this indication does net provide any control or trip functions. Stator temperature indicators are usually on generator units exposed to harsh environments. This is not applicable at CPS as the DG systems are located in environmentally controlled areas. -The Divisions I and 11 DGs do not have this indication. The only accidents evaluated in the USAR are in Section 15.2.6 which involves the loss of AC power. Not having stator temperature indication will not affect this accident. This change will not create a new type of accident as the temperature indication is not required for the safe operation of the DG systems.

- QA PROGRAM REVISIONS Document Evaluated: USAR Section 1.8, ORM 6.5.2.8 Log Number: 95-075 This change adds clarifications to USAR Section 1.8 for Regulatory Guide (RG) 1.33, "QA Program Requirements," and RG 1.144, " Auditing of QA Programs for Nuclear Power Plants."

1This changed the CPS commitment to perform audits in selected areas on the same frequency as stated in ANSI Standard N18.7 1976," Administrative Controls and Quality Assurance for the

' Operational Phase of Nuclear Power Plants." The procedure periodic review frequency requirements were chanpd. Audit frequencies were deleted from Operational Requirements Manual (ORM) Section 6.5.2.8._ This also changed the audit program to allow the auditing organization to determine root cause evaluation and follow-up. This change was determined to be a reduction it commitment and required prior NRC approval before implementation. This approval was provided and documented by NRC letter dated August 22,1995. These changes were reviewed under 10CFR 50.59 and the changes were administrative in nature and do not tre

' impact on the structures, systems and components (SSCs) at CPS.

CPS EMERGENCY PLAN AND EC-02 REVISIONS Document Evaluated: E-Plan, Rev 10 and EC-02, ACN 5/3 Log Number: 95-076-These chuges to Revision 10 of the CPS Emergency Plan (E-Plan) and EC-02, " Emergency Clarifications, do not decrease the effectivenes:: of the E-Plan per 10CFR50.54(q). These changes are mainly clarification and minor word changes. The changes in the E-Plan were for changing to the Improved Technical Specifications, organizational changes and minor changes to the Emergency Response Organization. These changes do not require revision to the USAR.

These changes do not affect the normal operation of any plant systems or equipment and, would

_ _ _ )

. . . . . _ . - - - . - . . - - - _ = _ - - - . - - - - - - .

} Attachment 2 '

, to U 602836

_ Page 16 of142 therefore, not involve an unreviewed safety question. The changes are mainly administrative in nature. _ _ ,

, L UPGRADE TO THE 3D-MONICORE SOFTWARE FOR CORE MONITORING Document Evaluated: Modification CX-30, Supplement 2 . Log Number: 95-077 ,

4 This modification supplement upgrades the core monitoring soRware from Honeywell computer based applications using the reactor heat balance soRwste to a star.J alone workstation in the 3 main control room using GE's 3D-Monicore soRware. The supplement deletes soRware no longer used. It adds to the Honeywell system revised soRware for processing Tracsing incore Probe (TIP)information. The supplement also removes the TRLMS PC used as a backup and p retires the magnetic tape drive which is no longer used. The soRware changes include the i _ deletion of the portiens of the Nuclear Steam Supply (NSS) functions no longer active and revises -

4 L the remaining functions to support the 3D-Monicore. This system is used to perform the nuclear

[ pertbrmance calculations for the core power distribution and thermal limits evaluation. This also affects the Operational Requirements Manual (ORM) because the 3D-Monicore soAware is able to calculate acceptable results with one TIP machine out of service indefmitely.

The uncertainty analysis in NEDE-32321, "3D-Monicore Evaluation Accuracy," determined that

_ the uncertainty of the new calculations is 5.4% for bundle powers and 7,6% for nodal powers.

These include the effects of an out of service TIP. This is compared to the uncertainty of 7.6%

using the former sonware. The criterion for these uncertainties is in USAR Section 15B.2 which i- 1 requires 8.7% to satisfy the uncertainties in the safety analysis. _ This modification had phased in i releases which allowed testing over a wide time span and potential plant conditions while always i-monitoring reactor conditions. By maintaining parameters within those specified in NEDE-32321, tM change would not decrease the margin of safety, create new accidents or increase the consequences of accidents evaluated in the USAR.

L

. CHANGE CONTAINMENT PENETRATION DESCRIPTIONS FOR IMC-169001 AND 1MC-079

l. Document Evaluated: USAR Table 6.2-47 Log Number:'95-078 This changes Table'6.2-47 to reflect that containment penetration IMC-169001 as a secondary containment bypass leakage path and that IMC-079 is not a containment leakage path. : Since ,

IMC-079 is not a containment atmosphere leakage path, for 30 days post-LOCA, the bnly 7

leakage past the isolation valves would be suppression pool water. This leakage will be included in with the hydrostatically tested valves. IMC-079 meets the requirements nf a closed system to

! preclude bypass leakage as described in Branch Technical Position (BTD) CSB 6-2. IMC169 (1

. - of 2) has no isolation provisions to prevent containment atmosphere leakage. The only effect on any assumptions in the USAR is on the calculation in Section 6.2.4.3.2.3 that calculates the effective total potential bypass leakage path value. This change does not change or add different

accidents previously evaluated in the USAR. In addition, this does not reduce margins or involve j a change to the CPS Technical Specifications (TS). Removing IMC-079 and adding IMC- ,

0 i

Attachment 2 to U 6028%

Page 17 of 142 l 169001 reduces the total bypass leakage rate from 832.8 cubic feet per day to 810.6.- This is L below the allowable TS limit of 908 cubic feet per day.

ALTERNATE DRYWELL FLOOR DRAIN ll.,0W MONITORING, SUPPLEMENT 1 Document Evaluated: Modification LD-027 S1 Log Number: 95-079 Modification LD-027, reported in Revision 6 to the CPS USAR (Log Number 95-056), installs a Leak Detection (LD) system flow element on the existing drywell floor drain system sump pump _

discharge line. The new element will provide a flow signal to a flow totalizer and will aid in determining the amount of unidentified operationalleakage. This new flow monitoring equipment wiil provide an alternate method for determining unidentified reactor coolant boundary leakage. The new equipment will provide a backup to the existing leak detection equipment and may prevent an unnecessarv plant shutdown should the leak detection equipment become inoperable. The USAR Revision 6 safety evaluation covered the first phase ofinstallation prior to releasing the equipment for opetation.

This safety evaluation covers the second phase of the installation. This phase added a programmable logic controller (PLC), a digital counter, a computer point, and two recorder channels in the main control room. The PLC receives input from the flow totalizers. The PLC calculates the average (total) and actual pump discharge flow by integ ating the signal from the flow totalizer converter and dividing the result by the total time between pump cycles. The second phase also changed this from an " alternate method for determining unidentified reactor coolant boundary leakage," to a means allowed to meet CPS Technical Specification (TS) 3.7.4.a, "RCS Lukage Detection Instrumentation." The CPS TS Bases Section 3.7.4 has been changed.

This alternate system will operate following an Operational Basis Earthquake (OBE) seismic event as the weir box system does.- This alternate system also meets the requirements of Regulatory Guide L45, " Reactor Coolant Pressure Boundary Leakage Detection Systen s," for accuracy and control room indication. No new failures are created by this change since the new equipment is passive in nature and the piping has been evaluated for the increased weight, i

RE-SPAN OF FLOW INDICATORS FOR HIGHER LEAKAGE FLOWS Document Evaluated: Modification LD-027 S2 Log Number: 95-080-This safety evaluation changed the parameters in the Modification LD-027 instrument channel from 50 gpm, contained within original phase one and two design documents (Log Numbers95-056 and 95-079), to 120 gpm. Supplement 2 (S2) of this design changes the span of flow sensor

. and the indicated flow range to O to 200 gpm, as well as makes changes to descriptions of pump flow and to description of an in-line flow meter. This new pump capacity is within the pipe design of the Floor Drain (RF) System. LD 027 S2 only changes the parameters in the instrumentation channel and documents the increased drywell sump pump flow. 'Therefore, there is no impact of

) this on the probability of any accident predously evaluated in the USAR. Margin of safety would -

not be impacted as the Leak Detec^ ion System is not credited for mitigating any of the design i

basis LOCAs in USAR Section 15.6.5, " Loss-of-Coolant Accidents."

Attachment 2 to U 602836 Page 18 of142 REFUEL BRIDOE CONTROL SYSTEM UPGRADE Document Evaluated: Modification FH-028 Log Number: 95-0G1 This n.odification upgrades the control system of the refueling bridge in the Containment Building. The upgrade installs a new Programmable Logic Controller (PLC) based control system and a personal computer to interface with the PLC. The design change also replaces the main hoist right and leR hand controllers, the position encoders, the main hoist load cells, and motor drive units. Additionally, the bridge control select switch will be relocated to mount directly on the bridge frame.

.. USAR Section 7.6.LI.3.3 describes the interlocks and the depth of protection they provide.

USAR Table 3.2-1 describes the refueling bridge as a seismic structure. The control system and

• components mounted on the bridge are not safety-related nor Class IE. This change does not affect the refueling bridge hoists, the grapple, or the drive motors. The bridge position switch and the all rods in interlocks are external to the PLC. Therefore, the installation of this chan8e will not cause a malfunction of equipment important to safety ~ USAR Section 15.7.4 evaluates a fuel handling accident. This change does not affect the grapple, or the hoist lifting and releasing interlocks. This change will not increase the probability of, the consequences of a fuel drop accident, or any accident previously evaluated. This change does not affect the interlock.-

. actuation points and thus the margin of safety is not reduced, ADD NOTE TO EECTION 4.6.5.1 OF THE OPERATIONAL REQUIREMENTS MANUAL Document Evaluated: ORM 2.6.5 Log Number: 95-083 This activity adds a note to the Testing Requirements Section which allows the travel stops to be temporarily removed from the Fuel Building crane to allow travel over the spent fuel pool provided the crane is unloaded and the main hook is positicned so it is not above irradiated fuel.

The main hook is rigged in accordance with single failure proof rigging procedures. USAR ,

Section 15.7.4, " Fuel Handling Accident," postulates drcpping of a raised fuel assembly onto stored fuel bundles. The most limiting scenario from a radiological point of view is dropping a fuel bundle on to stored fuel in the Fuel Building. The analysis of USAR Section 15.7.4 is bounding on this activity. The temporary removal of the travel stops and travel over the fuel pool in an unloaded condition will not increase the probability of a fuel drop accident or the consequences of such an accident. Temporary removal of the rail stops and the application of administrative controls to the fuel building crane does not ints; duce any new failure modes nor w change any operating parameters and will not increase the possibility of an accident not previously evaluated. Travel over the spent fuel pool in an unloaded condition with the main hoist positioned so as not to be over irradiated fuel does not increase the probability of failure of any equipment important to safety. This change does not change any setpoints, operating points, safety blassifications, or seismic qualifications for the fuel building crane. ~fhis change does not reduce the margin of safety.

Attachment 2 to U-602836 Page 19 of142 REPLACE SOLID RADWASTE SPENT RESIN TANK LEVEL INSTRUMENTATION

' Ikcument Evaluatedi Modification WX-028 - Log Number: 95-084 Modification WX-028 replaces the ultrasonic continuous level indicating system on the Solid Radwaste (WX) Spent Resin Tark OWXO4T with on-demand plumb bob level measuring instrumentation to measure both tctal tank level and sludge (interface) level. - The modification also installs a " Thermal Dispersion" RTD-type sensor element and circuitry for a continuous high-

- high level alarm. In addition, the 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 cannot be -

mnning due to the likelihood of damaging the plumb bcb. This modification is similar to WX-F010, reported in Revision 6 to the USAR. USAR Figure 11.4-3 Sheet 1, Appendix E Figures FP 16a&b, Appendix E Cable Tray Drawing 14, and USAR Section 7.7.1.12.3.1 are being changed by this modification. This modification is being made to improve the reliability oflevel monitoring instrumentation for this tank. The only related failure evaluated in the USAR is -

radwaste tank rupture. This change does not impact the probability or consequences of the radwaste tank rupture event.

CONTAINMENT REFUBL PLATFORM INTERLOCKS

. Document Evaluated: Temporary Modification 95-041 Log Number: 95-087 This activity temporarilyjumpers out three interlocks on the refueling platform in the Containment Building. The temporary modification (TM) was required to facilitate installation ofModification FH-028, which replaced the malfunctioning positioning computer and installed new position encoders. USAR Section 7,6.1.1.1 describes the interlocks as needed to prevent the reactor from _

becoming critical during refueling operations. USAR Section 7.6.1.1.3.7 requires functional

. testing of the interlocks prior to refueling operations. During the alignment of the position encoders, the mast will be stored and this TM will not increase the probability of malfunction of equipment important to safety. USAR Section 15.4.1.1 evaluates inadvertent criticality during refueling cperations. The TM was only installed during the implementation of Modification FH =

028 and fuel was not being moved. Therefore, the temporary modification did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. The temporary modification simulated conditions such that the three interlocks would allow movement over the fuel pools. The modification did not introduce any new movement or positioning of the refueling platform and as such, no new failure modes were introduced. The TM did not increase the possibility or the consequences of a new type of accident not previously analyzed. CPS technical specification applicability statements require the refueling interlocks to be operable during refueling operations, as this temporary modification was installed when the plant was not in a refueling mode, the margin of safety was not reduced.

Attachment 2 .

to U 602836 Page 20 of 142 -

ES CHECK VALVE DAILY EXERCISE Document Evaluated:t USAR Section 10.2.3.6 Log Number:: 95-088 This change modifies the USAR reouirements for a daily exercise of the extraction steam non-return check valves to a weekly exercise. The extraction steam non-return check vdves are classified as non-critical power assisted check valves. The power assist feature of these valves is nct required for over speed protection or water induction protection. The tu bine-generator is not required to support safe shutdown of the reactor or to mitigate the consequences of an accident. The turbine-generator could be a source of high energy missiles that could affect safety related structures, systems, or components. The primary protection against over-speed and water induction is the tuttine design itself. The check valves provide secondary assurance against over-speed or water induction. This change in exercise frequency will not impact that secondary protection. The change in exercise frequency from daily to weekly will not eliminate the testing requirement only modify it; therefore, this change will not increase the probability of failure or the consequences of failure of equipment important to safety. The change in escise frequency will

. not act as initiator for any of the accidents evaluated in the USAR, thus the change in exercise frequency will not increase the probability or the consequences of accidents previously evaluated.

This change does not introduce physical changes to any structure, system, or component and will

- not introduce any new failure mechanisms that may increase the possibility of occurrence or consequences of any accident previously evaluated in the USAR. Since these valves are not taken credit for in any Technical Specification Bases, the change in exercise frequency will not impact the margin ofsafety.

CONTROL ROD HYDRAULICS AND CONTROL Document Evaluated: CPS Procedure 3304.01 and Log Nurnber: 95-089 Valve Lineup 3304.0lV001 Procedure CPS procedure 3304.01 and valve lineup 3304.0lV001 were revised to improve the conduct of the tining surveillance. This procedural change was made in compliance with the requirements of Technical Specification (TS) Surveillance Requirement (SR) 3.3.2.1.9. The revision allows rod motion in special situations if the rod motion meets special requirements The

. Rod Pattern Control System (RPCS) Rod Withdrawal Limiter (RWL) is designed to limit rod withdrawal to prevent violating the Minimum Critical Power Ratio (MCPR) safety limit and the

- 1% plastic strain limit on the fuel cladding. Rod motion meeting the requirements of the SR, and limited by the RPCS RWL, will not invalidate any analysis nor reduce the margin of safety. This procedure revision allows continuous rod movement within the TS 3.3.2.1.9 requirements, and

.will not introduce any new failure mechanism that can act as an initiator for a new type of accident not analyzed in the SAR. Therefore, this procedure revision will not increase the probability of a failure or its consequences of a component important to safety. This change will allow rod motion under the restrictions and analysis of the technical specifications. It will not increase the probability of a reactivity addition accident evaluated in USAR Section 15.4.

-L Aunchment 2 F ' to U402836 Page 21 of 142 - -

4:

1 LPCS SYSTEM VENTING DUE TO PRESSURIZATION UPSTREAM OF INJECTION 4 iVALVE lE21-F005 -

Document Evaluated: USAR Table 6.3 11 Log Number: 95-090 h The change to USAR Table 6.3-11 allows periodic opening of existing drain valve IE21-F004 as

a means of venting pressure from the Low Pressure Core Spray (LPCS) System pump discharge piping. A condition report identified that the LPCS discharge lin.8, between the LPCS injection valve IE21 F005 and the LPCS pump discharge line check talve IE21-F003, was pressurizing to  ;

l a higher than expected value. This activity will allow venting of the LPCS discharge piping during a plant startup in order to preclude unnecessary lifting of the relief valve IE21-F018. The

venting activity will use the existing drain path, will require operator presence and

[  : communications with the Main Control Room, and will not introduce any new venting or *

-discharge path. The vent proces: will not introduce any air into the LPCS piping b9cause pressure is monitored to assure it does not fall below the waterleg pump discharge pressure.

Administrative controls require communications between the Main Control Room and the valve

- operator until the valve is re-closed and re-locked. The vent flow rate is low and well below 5  ;

E - gpm USAR Section 6.3.2.2.8 was reviewed to assess the impact on the LPCS system by the venting operations. The venting will not preclude the pump from starting and operating when

! required and as such will not increase the probability of failure of the pump. The venting

[ operation' will not act u an initiating event for any of the accidents previously evaluated in the i USAR. Therefore, the venting activity will not increase the probability of occurrence of or the i consequences previously evaluated in the USAR. The venting activity does not introduce or

!- modify any structures, systems or components and will only temporarily reposition the drain valve. Therefore, the venting activity will not increase the possibility of any r.ew accident or

malfunction not previously evaluated in the USAR. The venting activity does not affect any l acceptance criteria in the Technical Specifications and thus does not decrease the margin of r safety, FEED WATER HEATER LEVEL OPTIM1ZATION TEST / DATA SHEET 7

Document Evaluated

CPS Procedure 2800.69 Log Number: 95-091 and Data Sheets 2800,69D001-D011 4

The Feedwater Heater Level Optirr.ization Test (and the associated data sheets) is a new procedure that optimizes the operr. ting level and thermal performance of the feedwater heaters.

During the performance of the test the feedwater level will be varied in small increments to

= determine the optimum level. Temperature is not expected to increase or decrease by more than 6'F. The procedure will vary feedwater level, which will affect feedwater temperature.

g - Feedwster temperature will change reactivity, and reactor power. USAR Sections 15C.1.1.2,

. 15.1.1, and 15B.3.1 analyze power operations with feedwater heaters out of service, decreases in

~

reactor coolant. temperature due to loss of a feedwater heater, and failure of a feedwater heater e controller respectively. USAR Section 7.7.2.4.2 classifies the feedwater system as non-safety

[ related and not required for safe shutdown of the plant. Although in a worst case scenario the

[

procedure has the potential to cause tube damage if vibration or flow oscillations are induced, the i

~. -- - .- - -- -- - . ,-.

Attaciunent 2 to U-602836 Page 22 of142 procedure will_not increase the probability of malfunction of equipment important to safety. The procedure will vary feedwater level of the heater under test and temperature of the feedwater within a controlled margin to determine optimum feedwater heater level. Since, the resulting heater level will be within the design parameters it will not increase the probability of occurrence or the consequences of an accident previously evaluated. The procedure will determine an

= optimum operating level for the foodwater heaters, the final operating level will be within the feedwater heater isolation setpoints; and, if the final feedwater temperature varies by more than 6'F, the test will be terminated an 1 the plant returned to a pre-test candition.:Thus, the feedwater hester level optimization procedure will not introduce any new components or different operating points to increase the possibility of a new and unanalyzed accident. - The procedure will vary feedwater heater level and feedwater temperature, but will have no effect on the reactor protection setpoints and thus will not change the margin of safety.

ALLOW CARPET TILE INSTALLATION IN THE MAIN CONTROL ROOM COMPLEX, OSC & TSc.

Document Evaluated: ECN 29418 Log Number: 95-092 This change evaluated the addition of tile type carpet in the Main Control Room (MCR), the Operatiens Support Center (OSC), and the Technical Support Center (TSC). USAR Appendix E, .

" Fire Protection Evaluation Report" (FPER) was 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 carpet. The review included the requirements ofNUREG-0800, " Standard Review Plan," to detennine the effbets 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 wouki have no adverse effect on the equipment, instrumentation, or ventilation in the MCR, TSC, and OSC. Therefore, the probability of

. occurrence or consequences of failure of equipment important to safety would not incream Similarly, and for the above reasons, the addition of carpet would not increase the probability of an accident or its consequences previously evaluated, nor would the possibility of an accident or its consequences increase. Since the addition of carpet does not introduce any new parameter or operating characteristics, which could affect any setpoints or operating points, the addition of carpet does not affect the margin of safety.

At the time of USAR Rev. 7 this modiscation is not fully implemented._ A revised summary will be submitted with USAR Rev. 8.

INSTALL HIGH POINT VENTS ON THE RR DISCHARGE PIPING Document Evaluated: Modification RR-033 Log Number: 95-093 This Modification adds %-inch vent assemblies to the Reactor Recirculation (RR) Syrten' discharga piping 4-inch decon flanges. The vent assemblies consist of %-inch piping, Mach isolation valve, and a threaded pipe cap. The modification provides a means for venting the RR piping and to minimize the potential for introduction of particulates into the RR pump wals, thus i

l

Atuchment 2 -

to U 602836 -

Page 23 of142

_ decreasing the probability of RR pup seal failure. The vent assembly is attached to the Recirculation Loop and thus qualifies as high-energy piping. USAR Section 3.6.2.1 A exempts

- piping of 1-inch or less from being considered for pipe breaks. Therefore, the addition of the vent -

assemblies will not incruse the probability of mtifunction of or the consequences of a malfunction of equipment important to safety,- USAR Section 15.6.5 evaluates loss of coolant accident from

- postulated line breaks. The failure of the recirculation piping is evaluated as ont f these breaks.

The vent assemblies are designed and constructed to the requirements of the ASME boiler and pressure vessel code.- The piping analysis showed that the vent assemblies have no effect on the integrity of the RR loop piping. The probability of a RR line break is not increased. Similarly, the vent assemblies will not increase the consequences of a RR line break. The addition of vent -

assemblies will not increase the possibility of the design basis accident for the RR piping. There is no other credible new accident that could be created by the addition of the vent assemblies.

The addition of the vent assemblies does not affect the margin of safety as failure of the vent assembly is enveloped by normal water makeup capabilities.

CPS ORGANIZATION CHANGES Document Evaluated: USAR Chapters 12 and 13 Log Number: 95-094 Organizational changes at the Illinois Power Clinton Power Station require changes to Chapters 12 and 13 of the USAR. All described USAR positions continue to be filled by qualified personnel. These changes do not impact plant equipment, and as such, do not have the potential for increasing the probability of occurrence of any malfunction of equipment or any accident or the consequences of any accident analyzed in the USAR. These personnel changes have no effect on plant operating parameters and do not increase the possibility or consequer.ces of any accident

. not previously evaluated. Since this revision changes position titles only, it does not affect the margin of safety.

REVISE SUPPRESSION POOL MONITORING SYSTEM REQUIREMENTS Document Evaluated: ORM 2.2.15 Log Number: 95-095 This change adds new action statements to the Operational Requirements Manual (ORM). The change provides the actions necessary to allow operations with less than 16 suppression pool temperature monitors operable. Continued operation is allowed as long as at least one temperature monitor in each of eight sectors is determined to be operable. Additionally, the revision to the ORM removes the actior.2 associated with the suppression pool water level instrumentation. The suppression pool temperature instruments to which this ORM actions apply are not part of the Post Accident Monitoring Instrumentation of Technical Specification (TS) 3.3.3.1.

TS 3.6.2.1 does not specify that the suppression pool average temperature be deterinhied'using these instruments. The technical specification bases require that all operable instruments be used to di4 ermine the average temperature, yet do not specify the number ofinstruments that are required to determine the average temperature. The change also removes the action statements

Attachment 2 to U402836 Page 24 of142 J for the suppression pool water level instruments. The action statements removed by this change had no corresponding operational requirements or testing requirements in the ORM, and as such

= the action statement could not be entered. These changer do not affect the requirements placed on the instruments of TS 3,6.2,1 and 3.6.2.2. These instruments armt assum d operable in any accident analysis of the USAR. These changes will not physically change any structures, systems >

- or components. Therefore, changes will not increase the probability of failute of any component important to safety. These changes will not act as initiators for any accident sequence evaluated in the USAR and as such, w.:1 not increase the probability of occurrence or the consequences of any accident previcusly evaluated. These changes to the ORM action statements do not introduce any new failure mechanism and thus will not increase the possibility of or the consequences of any accident not previously evaluated. As these temperature and level instruments are not the same as

- those of TS 3.o.2.1 and 3.6.2.2, these changes will not decrease the margin of safety.

ADDITION OF FILTERS IN RD PIPING SUPPLYING WATER TO RR PUMPS Document Evaluated: ECN 29395 Log Number: 95-096 L This ECN adds filtration capability to the Control Rod Drive Water (RD) System supply to the Reactor Recirculation (RR) puraps. Two 100% capacity filters, associated piping, and l._ instrumentation are added to the RR pump seal purge supply line. Addition of filters in the seal purge water supply to the RR pumps enhances the quality of the seal purge water and reduces the potential for RR pump seal degradation. The potential for flow reductions or stoppages due to

clogged filters has been addressed by the installation ofinstrumentation, which monitors the differential pressure acrou the filters. Therefore, the probability of a malfunction of equinment l -important to safety is reduced by this change. USAR Section 15.3 evaluates a decrease in ne i reactor coolant system flow rate. Loss of seal purge water to the RR pumps would not cau>e a sudden loss of the RR pumps due to internal seal flow and external cooling provided by the

Component Cooling Water system. Thus, this change does not increase the probability _of

l. - occurrence or consequences of an accident previously evaluated in the Safety Analysis.- The

- analysis for loss of one recirculation pump is in USAR Section 15.3. Technical specifications allow operation with one recirculation pump. Trip of both recirculation pumps is also analyzed

- and reactor protective actions occur, thus, the margin of safety is not reduced by this change.

ACTUATOR GEAR CHANGES FOR EIN'S 1E32-F006 AND 1E32 F009 Document Evaluated: ECNs 29241 and 29242 LogNumber: 95-097 During reactor depressurization, any vapor leakage past the Outboard Main Steam Isolation

, Valves (MSIVs) IB21-F028A, B, C or D will flow through the Outboard MSIV Leakage Control l System (LSC) valves IE32-F008 and IE32-F009 and be discharged to the steam tunnel area. In l the bleed mode of operation, any vapor leakage past the Outboard MSIVs will process through the MSIV LCS bleed valves IE32-F006 and IE32-F007 to mix with dilution air from the

! atmosphere. This dilution air is extracted from the Auxiliary Building by the MSIV LCS outboard j- air blower IE32-C002F. These air blowers will discharge the final mixture to the inlet of the l Standby Gas Treatment (SGTS) System for processing.

l I.

[.

I I

i Attachment 2 to U 602836 Page 25 of142 The modification changes the gear ratios for these valves to increase the torque. This change

, slowed valve closure rates. - As such, USAR Section 6.7.2.3.b, " Valves," lists valve stem speed ,

! - requirements of, "11 in/ min for the gata valves and 4 in/ min for the globe valves." A note is being '

3 added to this section stating that the speed requirements above are not applicable to IE32 F006 and IE32-F009 and that any future changes to other MOW in the MSIV LCS will be evaluated 3- on a case-by-case basis. Once modified, the valves will still continue to operate and function to j meet all requirements of the GE Design Specification 22A4674AD Revision 8 and S&L Design -

L - Specification DS-IS-01-CP. The valves are not Containment Isolation Valves (CIVs). Therefore, l= closure speed would not be an issue for 10CFR100 offsite dose; no increase in consequences of F an accident. The MSIV-LCS can be manually actuated after a LOCA har occurred, provided that the reactor and steam line pressure are below the pressure permissive interlock set points and the inbosid MSIVs are fully closed. Per USAR Section 15.6.5.5.1.2, " Assuming the MSIVs leak 28 CFH per valve, leakage past the inboard MSIVs is estimated to begin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the accident."

Increasing the stroke time of this valve from 38 to 48 seconds for 2 inch lines is inconsequential in

. this timeline of events. This change has not been fully complete as of the cutoff date for USAR Revision 7, However, portions have been installed. An updated status of this change will be 4

reported in Revision 8 to the USAR.

E' HEADQUARTERS SUPPORT CENTER OPERATIONS AND STAFFING

~

' Docurr,ent Evaluated: EPIP HQ-01 Log Number: 95-099 I Emergency Pihn Implementing Procedure (EPIP) HQ-01, " Headquarters Support Center

Operation and Staffing," is changed to update titles, and incorporate comments generated during

- recent drills. This change is administrative in nature and does not affect any plant equipment,

-operating parameters, or setpoints. The procedural change will not increase the probability of occurrence of any equipment important to safety or of any accident or its consequences

previously evaluated in the USAR. Similarly, this change will not increase the possibility of any l nr.w type accident, or its consequences, not previously evaluated. These are administrative changes and would therefore have no impact on the margia of safety as described in the technical

- specification bases.

ROD BLOCK REQUIREMENTS CHANGES IN ORM L

Document Evaluated: ORM 2.2.1,2.2.2,2.2.3,2.2.4 Log Number: % -002 r

i This change eliminated the Operational Requirements Manual (ORM) requirements for rod blocks

. originating from the following instrumentation during MODE 5: Source Range Monitoring

(! - (SRM), Intermediate Range Monitoring (IRM), Average Power Range Monitoring (APRM) and

, Scrua Discharge Volume (SDV) level instrumentation. This requirement was eliminated so -

surveillance tests associated with this instrumentation would not be required to be performed I while the plant is in MODE 5. This change only impacts the operability during MODE 5.

Technical Specification (TS) 3.3.1.2 requires that sufficient indication from the SRMs be operable

to adequately monitor the neutron flux level in the core in MODE 5. This change does not impact

that operability requirement. USAR Section 15.4.1.1, " Control Rod Removal Error During

Refueling," does not take credit for these level control rod blocks. Therefore, this change would l

Attachment 2 to U402836 Page 26 of 142 not irnpact design basis accidents. Shutdown margin inadvertent criticality is also ensured via TS 3.1.1,"SilUTDOWN MARGIN (SDM)." TS 3.9.1," Refueling Equipment Interlocks," ensures that no control rod can be withdrawn when the refuel platform la over the reactor core. These requircinents ensure that the change has no adverse impacts on safe operation and the margin of safety for reactivity accident analysis.

COMMUNICATIONS .. REFUELING OPERATIONS, ORM 3

Document Evalua 1. ORM 4.6.2 Log Number: 96-003 This chrnge removed the requirement that direct communications be demonstrated one hour prior to the start of CORE ALTERATIONS. Ilowever, the requirement to demonstrate direct communication at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during alterations has been maintained. This change will allow the Operational Requirements Manual (ORM) testing requirement to be performed on the same frequency as other surveillance's performed during refueling operations. ORM Section 5.6.2 is also being revised for consistency in referring to refuel platform personnel. The USAR does not identify that this action is required for mitigation of accidents. This change does not chminate the requirement to maintain oirect communication between the Main Control Room (MCR) and the refuel phtform. Therefore, no new failures would be caused by this change.

Removing this unnecessary test does Sol affect protective barriers, thereby massin, as this system does not provide an active safety function or feature. USAR Section 9.5.2 describes the design and operation of the plant communications system. The system is non safety related and as such, would not impact cny equipment or malfunctions as described in the USAR.

OFF-SITE DOSE CALCULATION MANUAL (ODCM), REVISION 13 Document Evaluated: ODCM, R13 Log Number: 96-004 Changes were made in the Off Site Dose Calculational Manual (ODCM) to correct typographical errors; provide clarMcation to existing ODCM requirements for sampling requirements and Lower Level of Eavetion (LLD); standardize the definition of: Channel Calibration, Function Checks, Operabiky, Remedial Requirem:nts and Surveillance Requirements between the improved Technical Specifications (ITS) and ODCM. The Erwironmental Report-Operating License Stage (ER-OLS) Table 5.1-4 has been deleted as the table was required for the pre-operational phase of CPS and is not longer required. Appropriate r aperational sample points were incorporated into the operational phase (ODCM Table 51 ' additionally, Licensing Amendment request #LC 69 was deleted as the amendmF w aroved. These changes do not affect how equipment is operatul at CPS These changes are r ly administrative in natore.

Therefore, this change would not involve changes to acciderm .dfunctions or margin of safety as described in the US AR.

t k

un _ - _ _ _ _ _ _ - \

f Anachuneet 2 to U 602sM Page 27 of142 RADIOLOGICf PROTECTION PROGRAM CHANGES Document Evaluated: USAR Sections 9.4,12.1,12.2,12.3 & 12.5 LogNumber: M-005 These changes provide for a decision point by Radiation Protection Supervision to determine if persing the Reactor Pressure Vessel (RPV) through the drywell purge system is warranted. This is changed in the sections related to refbeting (USAR Sections 9.4.7.2.1.2.h, 9.4.7.2.2.e,12.2.2,4 and 12.5.3.2.1). During the actual RPV dissanembly process, the gas _ void in the vessel is collapsed by Alling the RPV to just below the Aange. Connecting the purge line fkom the RPV vont to the ventilation system is both time consuming and dose intensive. As such, this should only be performed when necessary.

Changes were made to more accurately renect the program description in Sections 12.1.3, 12.3.1.6,12.3.2.4.a and 12.5.3.2.6. These changes were to the description of the Radiological Protection Program audit (specific topic) versus the ALARA Program audit (general topic);

Identined Radwaste processing (general) versus drumming (specific) and identifies the

- radiological class!fication of risk which requires a Speci6c Radiation Work Permit.

Section 12.3,1.7.4 was revised to acknowledge the CP-020 modification being installed This was evaluated with Safety Evaluation Log Number 94 075 Rt.' These changes do not involve any ci.anges to the operation of equipment important to nuclear safety. Therefore, these changes do not affect accidents or malfbnctions of equipment as described in the USAR. The change. do not impact any NRC acceptance criteria (i.e.,10CFR19.12,10CFR20); therefore these chariges would not reduce margin of safety.

REMOVE MONTHLY DG AIR DRYER DEW POINT CHECK REQUIREMENT Document Evaluated: USAR Section 9.5,6.2 LogNumber: 96-006 This change removes the requirement 'o perform monthly dew point checks for the Divbion I,11 and 111 Diesel Gerprators (DGs). This requirement will be changed to, " periodically." This will allow the periodicity of preventive maintenance (PM) tasks to be performed based on the results of past PMs. The dew point acceptance limits of the air dryer system is 20 'F. 'Ihe average dew

. point for the two years prior to this change was negative 30 'F with no test above the acceptance limit of 20 'F, The initial periodicity for this PM will be quarterly. This is the vendor's recommended periodicity. However, if quarterly testing would show adverse trends, the

' periodicity may be adjusted based on the results of the dryer checks. The required dew point of 20 'F as stated in USAR will still be met. Any adverse trend would still be found. Therefore, this change would not impact the reliability of the DG system or increase the frequency of equipment malfunction. Since the DG reliability will be unaffected, barrier integrity will be maintained along with the margin of safety.

= -

Attaciument 2 to U4028%

Page 28 o(142 SHIFT TECHNICAL ADVISOR (STA) TRAINING PROGRAM Document Evaluated: USAR Sections 13.1,13.2 and Appendix D LogNumber: 5007 These clar.nges revise the Illinois Power (IP) commitment to the ShlR Technical Advisor (STA) program in the USAR; The changes are based on recent analysis and are consistent with 10CFR50.120, " Training and QualiAcation of Nu; lear Power Plant Personnel," which dennes the

- reqairements of the STA program. The method of developing the STA training program content

- has changed firom one prescribed by the NRC in the original program to one developed by the utility using a systems approach to training (5AT). This program uses guidance of the National Academy for Nuclear Training document ACAD 91-016, "W Process for Accreditation for Nuclear Training in the Nuclear Power Industry." As an INPO. accredited program, the CPS ST A training program meets the req'tirements of 10CFR50.120 and the 10CFR55A definition of "sykJa approach to training." Ths STA training program is incorporated by reference to IP's commitment to Regulatory Guide 1.8 in USAR Section 1.8. As a change to training, no equipment is directly affected by the change. This program will continue to meet the

, requirements of the regulations. This training addresses off normal and accident situations using -

approved procedures to mitigate accident and transient scenarios. As an " advisor" to the shiA supervisor, the STA does not have line authority to direct shiR activities or the authority to deviate from approved procedures. Therefore, the change would not increase the likelihood or consequence of an accident or malfbnction.

MSIV SOLENOID VALVE REPLACEMENT Document Evaluated: Modification MS F031 Log Number: % 008 This modification provides a pair of automatic valve single coil solenciid operated valves (SOVs) as a substitute fnr the discontinued ASCO dual coil nuclear SOY (ASCO 6eries NP8323). These valves control pilot air on the Main Steam isolation Valve (MSIV) actuators. The mountms, air tubing and raceway are ofIllinois Power (IP) design. The design incorporates an EGS brand multi pin connector at the conduit entrance to the MSIV electrical box =. This is a new environmental seal to the conduit. This is required due to the difference in qualification from the ASCO to the new SOVs.

The two new 3 way SOVs have been tubed to provide the same function as the old dual coil SOV6.

The t.ibing schematic and the two in*vidual solenoid coilt. 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.

USAR Figure 3.6-1 shows the main steam lines as high energy piping which is evaluated for high energy line breaks, This modification maintains the same code qualifications for the MSIV. 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 USAR. . As this modification replaces porting internal to the dual coil SOV body with external tubing, it would not create the possibility of equipment failures different from those already analyzed in the USAR. Since this does not afrect closure times or induce teactor internal transients, the margin of safety will not be impacted. This change has not been fully

- - = = _ = = . _

Auachment 2 to U 602:36 Page 29 of142 complete as of the cutotdate for US AR Revision 7. However, portions have been installed. An updatei status of this change will be reported in Revision 8 to the US AR.

OPERATIONS CONTINUING TRAINING PROGRAM REVISION Document Evaluated: USAR Sections 1.8 and 13.2 LogNumber: %-009 These changes revise the Illinois Power (IP) commitment to the Operator Initial and Continuing Training Programs. USAR Section 1.8 lists the commitment to Regulatory Guide 1.8, Proposed Revision 2, " Personnel Selection and Training," which endorses list specific training program requirements for initial and continuing training. The CPS Licensed Operator Initial and Continuing Training Programs are accredited by the INPO National Academy for Nuclear Training. The programs are based on a systems approach to training (STA) using guidance of the National Academy for Nuclear Training document ACAD 91-016,"The Process for Accreditation for Nuclear Training in the Nuclear Power Industry." The CPS training programs meet the requirements of 10CFR55, " Operator's Licenses." Exception 7 to Regulatory Guide 1.8 in USAR Section 1.8 la being revised to reflect these changes as Exception 8 is being deleted to make the USAR consistent with current regulations in 10CFR55, 10CFR55 allows the program content to be developed using a systems approach to training. As this is a change to the training description, no equipment is directly efected by the revision. This program will continue to meet the requirements of the regulations. Since thl change meets regulatory requirements and implements INPO accredited guidance, the change would not increase the likelihood of an accident, malfunction or decrease in ma. gin.

INCLINED FUEL TRANSFER SYSTEM (IFTS) - REFUELING OPERATIONS Document Evaluated: ORM 2.6.6 LogNumber: %-010 These changes include dividing the equipment associated with inclined Fuel Transfer System (IFTS) into two categories; that which is required wherever IFTS is in operation not transporting irradiated components and second, those requirements which are required when IFTS is being used to transport irradiated components. This change la being made so that sections concerning radiological safety to personnel are required only when radiological hazards are present. The applicability statement was changed to more clearly state when IFTS is required to be operable, The testing requirements to verify personnel are out of areas adjacent to IFTS and doors are locked is being changed from one to four hours prior to IFTS startup. Testing of the IFTS is being revised from seven to 14 days based on an evaluation of equipment design and failure data and equipment significance US AR Section 9.1 A.5.3 describes the IFTS interlock restricting area access affected by IFTS transfers. This change relaxes this requirement such that when transferring items that are not irradiated, such as new fuel or during pre-outage testing, the interlocks would not be required to be in place.

This change will not affect the commitment to 10CFR20 limits. Areas and doors having the IFTS access interlocks are posted " Lethal radiation dose" on the doors. Therefore, it is unlikely that individuals would unknowingly enter these areas, The system design, reliability and administrative controls offset any potentialincreased consequences of accidents involved with this change. This

- = =__ _

Attachment 2 to U 602836 Page 30 of142 system has been used and mon'tored during the last four refueling outages (covering 86 total usages) without a single failunn. The worst can fuel handling accident does not involve or teke credit for the IFTS. This systom is not relied upon or used to prevent or mitigate the consequences of any accidents evaluated as design basis.

REMOVAL OF CONSOLE P679 FOR THE MCR HORSE SHOE AREA Document Evaluated: ECN 29432 Log Number: 96 011 ECN 29432 removes the Supervisory Monitoring Console IH13 P679 fkom the horse shoe area of the Main Control Room (MCR) and covers the resulting floor opening with a phenolic floor plate, A floor termination box will also be installed in the new floor plate to provide electrical power and communication ca~ ole terminations. This will permit future connections to a computer at this location. P679 displays plant parameters which can be monitored during normal plant operation, start-up, transients, or surveillances. This conr. ole was odginally placed at this location so the shift supervisor would have easy access to the display, Open access to thl console has been blocked by panel P800-64C. USAR Section 7,5,1,1.6 which describes this console as a means for supervisory personnel to observe plant operation without disturbing the operators will be deleted. _ Operations personnel indicate that other Cathode Ray Tubes (CRTs) inside the horse shoe area or those in the Computer Room would be used instead without adding additional burden. The CPS simulator was modified to reflect this change. During a four week period, requalification training was conducted with the ! 579 removed. This did not cause concerns to the shift supervisors or operators during this tralrdng.

The change was evaluated for fire protection impact and fire load changes were accounted for in Illinois Power (IP) calculation IP-M-0177, R/3, VolumeJ. This change did not require USAR changes for fire load impact to the USAR. No control function is available at the console. Floor loading was reviewed; previous analyses envelope these changes. Electrical loading was reviewed; previous analyses envelope these changes. Therefore, this change would not create new equipment malfunctions or accidents. This change does not impact the Technical Specification Bases nor does it change the margin of safety.

DEPTH OF ACTIVE FUEL BELOW POOL SURFACE DURING REFUELINO TRANSFERS Document Evaluated: USAR Sections 9.1.4.2.7.1,9,1,4.3, ORM 4.6.3.c - Log Number: % 012

- This changes USAR Sections 9,1.4.2.7,1 and 9.1.4.3 and Operational Requirements Manual (ORM) 4.5.3.c to provide consistent verbiage and avoid confusion. These changes are to say that, "The grapple in its Normal Up interlock position provides 8 feet 6 inches minimum water shielding over the active fuel during transient." The ORM is changed from " irradiated" fuel to

" active fuel" for consistency. This change is being made to avoid confusion with the mast's storage position which is 16 inches above the Normal Up position. This change will better describe this system and its operation. By clarifying the Normal Up interlock position, it helps prevent movement of active fuel to locations where it would have less than required water shielding coverage. Fuel handling accidents are evaluated in USAR Section 15,7.4. The fuel handling accident is assumed to occur as a consequence of a failure of the fuel handling assembly

h hhh -

s .

Attachment 2 l to U 602836 Page 31 of 142 lifting mechanism. This change does nM Npet tb .dem. As such, this change does r. st affect refueling operations or the probability of refueling accidents. If radiation levels were above 10 mrem per hour, area radiation monitors located on the refueling bridge would cause a trip to f stop the holsts upward motion. Therefore,10CFR20 limits would still be maintained and active fi cl barriers / margin of safety would not be negatively impacted.

CONTAINh!ENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEh!

1NSTRUhiENTATION Document Evaluated: ORh12.2.16 Log Number: 96-013 This change revises Operational Requirements hianual(ORhi) Section 2.2.16 requirements for containment and reactor vesselisolation control system instrumentation. This change deletes Note "d" which states, "Also trips and isolates mechanical vacuum pumps", in ORhi Table 3.2.161 and incorporates its :equirements into a new item on the same table. This change is being made to make it clear that the trip and isolation function of the mechanical vacuum pumps on a main steam line high radiation signal is required to be operable. The isolation function of the mechanical vacuum pumps from the condenser is actually caused by the pump tripping on a high main steam line radiation signal not on the high radiation itself. The failure to understand the trip and isolation of the mechanical vacuum pumps operation resulted in failure to test this trip function as required by the ORhi(LER 96-01). Also, the applicability of the trip and isolation of the mechanical pumps is being changed to be required in hiODES 1,2 and 3 when the mechanical vacuum pumps are operating or are not isolated. This change does not affect equipment operation. Therefore, the probability of equipment malfunction would not be changed. The design bases analysis for the Control Rod Drop Accident assumes a 1 percent per day condenser leakage. This would not be impacted since the trip and isolation of the mechanical vacuum pumps is still required. This change does not affect any protective barrier or the Technical Specification Bases. The trip and isolation function of the pumps are still maintained in hiODES 1,2 and 3.

REFUEL PLATFORhi UPGRADE TEST PROCEDURE FOR hiODIFICATION Fil-028 Document Evaluated: CPS Procedure 2800.62 LogNumber: 96-014 CPS procedure 2800.62, " Refuel Platform Upgrade Test Procedure (FH-028)," is a new test procedure for hiodification Fil.028 and its supplements. This modification upgrades the control system for the refueling platform in the Containment Building. This procedure will be used to verify the correct operation of the equipment on the refueling pistform: the new bridge motor controls, trolley motor controls, hoist motor controls, grapple controls, load cells, all the interlocks, boundary zones, the semi automatic feature, the frame and monorail hoists. The procedure calls for adjustment of and verifies interlocks to ensure the refuel bridge is operating properly. All the testing aethities for this will be conducted before the modification is released for operations. Therefore, this will r.ot impact equipment important to safety. Fuel handling accidents are evaluated in USA.R Section 15.7.4. The Fuel Handling Accident is assumed to occur as a consequence of a failure of the fuel handling assembly lifting mechanism. Since this test does not move actual fuel bundles, it cannot increase the probability or consequences of an accident evaluated in the USAR. The non-refueling outage portion of the test will temporarily

~ _ __ _ __ ____ _ _ _ _ __. _ ._ _ _ _ _ __ _

Anaciument 2 m U.602sw Page 32 of142 i - override the refbeling interlocks. However, this feature is only required during in vessel fbei 4 moverments. W refbeling outage ponion of this test does not override these interlocks.

l UNDERVCi.TAGE PROTECTION UPGRADE Document Evaluated: Modification AP-028- Log Number: 96-015 This uxilfication will replace the existing second level undervoltage relays with more accurate  !

ones. It will also replace the existing Motor Control Center (MCC) distribution panel transformers l with regulating ones and change the tap settings on lHG03JB. The goal of this modification is to anrure that under LOCA loading conditions, there is sufficient voltage at all the engineered safety

features (ESP) equipment. 'This modification will bring CPS into compliance with the requirements i of10CFR50 Appendix A, GDC 17 and assure that there is adequate voltage from the off site L power system. The modification meets the requirements of the Standard Review Plan and the NRC j BTP PSB 1 contained in NUREG-0800, " Standard Review Plan for the Review of Safety Analysis l Reports for Nuclear Power Plants." This modification is consistent with the CPS response and l 1 commitments made in the Illinois Powe. (IP) response to USAR Q&R 430.135 and documented in i the NRC Safety Evaluation Report (SER) NUREG-0853, Supplement 7, " Safety Evaluation Report -

L Related to the Operation of Clinton Power Station, Unit No.1."

i- The equipment used in this modification will not increase the probability of a malfunction of .

equipment important to safety. The new relays and regulating transformers are qualified IE per  !

j IEEE 323 1974, " Standard for Qualifying Class IE Equipment for Nuclear Power Generating

?

Stations," ar.d 344-1975, " Recommended Practices for Seismic Qualification of Class IE l

. Equipment for Nuclear Power Generation Stations." h failure modes have not changed except for the potential overvoltage which could be a failure mode in the regulating transformer.

Therefore, the regulating transformers are equipped with an overvoltage device. This etTect would be the same as if the transformer failed open, a previously-analyzed scenario. The consequences of a malfunction are the same as previously analyzed because the failure modes are the same as analyzed in the original design. This change does require a Technical Specification (TS) change for degraded voltage relay settings. This change was approved by the NRC on December 4,1996. Therefore, the TS change has been evaluated. No change in TS Bases or margin of safety as defined in the Bases had been determined. This issue was recognized and -

reported in LER 94-005 date June 3,1994.

- WX TANK LEVEL MODIFICATION - UNIT 2 PHASE SEPARATOR TANK (2WXO3T)

Document Evaluated: Modification WX-029 Log Number: %-017 Modification WX-029 replaces the ultrasonic continuous level indicating system on the Solid Radwaste Processing (WX) tank 2WX03T with on-demand plumb bob level measurini instrumentation to measure both total tank level and sludge (interface) level. The modiScation also installs a "hrmal Dispersion" RTD type sensor element and circuitry for a continuous high-high level alarm. In addition, the pump trip and high and low sludge level alarms are being eliminated due to the manual rrsture of the new level monitoring system, The pump cannot be running due to the likelihood ofit damaging the plumb bob. This modification is similar to WX-s ha--%--.,--43--+-- w - ,,w,,, - - --wwwrr. w re e ,w w + v-= .---rw-,,=--eseww"- t -m ser-'ir--='-==r-vr wr ege e--ee v -*vv ve*+4 + vw'-rvv F v 'r=r "er- 9wM'y*w -rur --~ wee'sNw w% PT

Anachment 2 to U 602836 Page 33 of 142 F010, reported in Revision 6 to the USAR. US AR Figure 11,4 3, Sheet 5, Appendix E Figures FP 16a A b, Appendix E Cable Tray Drawing 14, and Sections 7.7.1.12.3.1 and 1.8 for Regulatory Guide 1.143, " Design Guidance for Radioactive Waste Management Systems, Structures, and Components installed in Light Water Cooled Nuclear Power Plants," are being changed by this modification. This modification is being made to improve the reliability of the level monitoring instrumentation for this tank. The only associated safety related failure evaluated in the USAR is WX tank rupture. This change does not impact the probability or consequences -

of the WX rupture event.

ELIMINATE RELIANCE ON THERMO LAG IF FIRE AREA CB-4 AND FIRE ZONE CB-Sa Document Evaluated: Modification FP-091 Log Number: %-Oi9 Fire area CB-4 is the Division I cable spreading room. For a fire in this area, safe shutdown method 2, using Division 11 equipment is utilized to achieve safe shutdown. Three method 2 safe shutdown cables (IRP02C, IRP02H and IVX28E) are routed through this area and wrapped with Thermo-Lag to prevent fire damage. Fire 1.one CB Sa is the Division 111 switchgear room.

For a fire in this area, safe shutdown method I, using Division I equipment is utilized to achieve safe shutdown. Two method I safe shutdown cables (IRP01C and 1RP0lH) are routed through this area and wrapped with Thermo Lag to prevent fire damage. Three method 2 safe shutdown cables (IRP02C, IRP02H and IVX28E) are also routed through this area and wrapped with Thermo-Lag to prevent fire damage. Cables IRP02C and 1RP02H are the DC ar,d AC power feeds for the Division 11 NSPS inverter, Cable IVX28E is the power feed for the Division 11 inverter room cooler, Modification FP-091 assigns new cables and re-routes power feeds so that they do not pass through either fire area CB-4 or fire zone CB $a. This eliminates the need to rely on Thermo-Lag to protect method 2 cables in CB-4. It also allows method 2 to be utilized as the safe shutdown method for fire zone CB 5a which eliminates the need to rely upon Thermo-Lag to protect method I cables in the fire zone, This change affects the routing of the cables. The equipment and power sources are unchanged.

The same wire will be used. Configuration changes have been structurally and seismically analyzed. Fire hazards analysis also has been addressed using NUREO-0800, " Standard Review Plan," Section 9.51, " Fire Prevention Program," guidance. As such, cable rerouting does not create or change equipment or accident scenarios evaluated in the USAR. Only the equipment credited as operable and in service during and after a fire in CB-4 and CB-Sa is changed.

Therefore, margin of safety would not be impacted by this change.

MAINTAIN VALVES IB21F016 AND IB21F019 IN A CLOSED POSITION Document Evaluated: CR l-%-02-040, Caution Tag 96 396 LogNumber: %-020 Due to excessive packing leak on valve IB21F016 (main steam line inboard drain isolation valve),

it is being maintained in the closed position. Valve IB21F019 (main steam line outboard drain isolation valve), is being maintained in the closed position for the same reason. USAR Table 6.2-47, " Isolation Valve Summary for Lines Penetrating Containment," lists these valves as "normally open" valves. As such, a safety evaluation was performed to evaluate the effects on plant safety

i Attachment 2 I to U402:36  !

Page 34 of 142 l

i p

while operating in this configuration. The valves are being maintained in the closed position

during power operation, but are not being disabled. These valves could fulfill the power generation design functions using the CPS Emergency Operating Procedures (EOPs). Steps exist

! within EOP 4411.09, "EOP RPV Pressure Control Sources," to cycle these valves when required.

The design basis function of these valves is to isolate the containment in the event of accidents. In order to achieve this ihnction the valves must close. Tagging these valves in the closed position j will keep the valves in their design basis ibnction position. When an isolation signal is received by these valves, no change from the already closed position would occur. As such, this would not l increase the consequences of an accident, increase off-site doses, or decrease the margin of safety.

{ VALVE REPLACEMENT FOR 1E12 F024B i

Document Evaluated: Modineation RH 045 Log Number: %-021 i

Modification RH-045 replaces valve IE12 F024B. The Losv Pressure Coolant Injection (LPCI) l B System injection valve, IE12 F024B, is part of the Residual Heat Removal (RH) System. The

! existing 14 inch 300 lb. class flex wedge gate valve will be replaced with a 14 X10 X14 reduced port 300 lb. flex wedge gate valve. Additionally, a % inch hole cill be drilled on the RH pump side of the disc face; this will eliminate potential pressure locking t oncems. The disc, disc guides, I and seat rings will be hard surfaced with a cobalt chrome alloy to rcJuce internal friction and l~ reduce operating thrust requirements by approximately 50 %. Additional changes to optimize the

! valve / actuator capability and accommodate the new valve stem will be made. The modification is l required because Generic Letter 89-10," Safety Related Motor-Operated Valve Testing and Surveillance," dynamic VOTES testing showed the existing valve is d: graded and has an j abnormally high valve factor. Although the present valve is operable it has very little operating i

and switch margin settings. Further degradation may make it impossible to set the valve torque )

l switch within acceptable limits. The replacement valve will have a shorter stroke time than the L old valve, and will eliminate the requirement to declare LPCI "B" inoperable when the valve is

l. open for surveillance testing or suppression pool cooling. J V .

The design, fabrication, and installation of the replacement valve will meet the same quality levels ,

i and ASME standards as the original valve and as the RH system. Since the operation and design i

of the replacement valve is the same as for the old valve neither the probability of failure of i > equipment important to safety nor the consequences of failure increase. The RH functions fc,r  !

i

. lE12-F024B and affected by this modification are RH LPCI "B" injection, RH Loop "B" J

suppression pool cooling, and containment isolation. The improved stroke time will allow IE12-  !

, - F024B to be open for surveillance testing or suppression pool cooling without declaring LPCI I

l "B" inoperable. For containment isolation, the improved stroke time will improve the isolation i function. 'Ihe % inch hole in the disc face will not result in unacceptable leakage because the  :

maximum accident containment pressure of 9 psig is insufficient to impact leakage since L containment isolation is provided by the upstream disc face (containment side of valve). For the

!- suppression pool cooling function the valve will have a slightly larger pressure drop; however, the

! system flowrate is not affected because the pump has adequate margin to maintain the flow rate. i l Therefore, the replacement valve will not increase the probability of occurrence nor the

consequences of an accident evaluated in the accident analysis. The replacement valve will not i . change the operation of the system, will not introduce any new failure modes, nor change system l

l l

' __:_.,-.- , -._ . _ _ _ _ _ , _ _ _ ~ ~ _ - _ , -___.,._-._,_,__,,,,-__m - _ _ . , _ . -

Anachmou 2 to U402836 Page 35 of142 operational parameters. Therefore, the replacement valve will not increase the possibility or consequences of an accident not previously evaluated. The replacement valve will not change RHR performance criteria defined in the USAR or in the Technical Specifications, therefore the margin of safety will not be decreased.

VALVE REPLACEMENT FOR 1E12 F024A Document Evaluated: Modification RH-046 Log Number: % -022 Modification RH 046 replaces valve IE12 F024A. The Low Pressure Coolant Injection (LPCI) A System lisection valve, IE12-F024A, is part of the Residual Heat Removal (RH) System. The existing 14 inch 300 lb. class flex wedge gate valve will be replaced with a 14 X10 X14 reduced port 300 lb. flex wedge gate valve. Additionally, a % inch hole will be drilled on the RHR pump side of the disc face; this will eliminate potential pressure locking concerns. The disc, disc guides, and seat rings are hard-surfaced with a cobalt chrome alloy to reduce internal friction and reduce operating thrust requirements by approximately 50 %. Additional changes to optimize the valve /

actuator capability an.d accommodate the new valve stem will be made. The modification is required because Generic Letter 89-10, " Safety Related Motor-Operated Valve Testing and Surveillance," dynamic VOTES testing showed the existing valve is degraded and has an abnormally high valve factor. Although the present valve is operable it has very little operating and switch margin settings, further degradation may make it impossible to set the valve torque switch within acceptable limits. The replacement valve will have a shorter stroke time than the old valve, and will eliminate the requirement to declare LPCI "A" inoperable when the valve is open for surveillance testing or suppression pool cooling.

The design, fabrication, and installation of the replacement valve will meet the same quality levels and ASME standards as the original valve and as the RH system. Since the operation and design of the replacement valve is the same as for the old valve neither the probability of failure of equipment important to safety nor the consequences of failure increase. The RH functions for IE12-F024A which are affected by this modification are RH LPCI "A" injection, RH Loop "A" suppression pool cooling, and containment isolation. The improved stroke time will allow IE12 F024A to be open for surveillance testing or suppression pool cooling without declaring LPCI "A" inoperable. For containment isolation, the improved stroke time will improve the isolation function. The % inch hole in the disc face will not result in unacceptable leakage because the maximum accident containment pressure of 9 psig is insufficient to impact leakage since

- containment isolation is provided by the upstream disc face (containment side of valve). For the suppression pool cooling function the valve will have a slightly larger pressure drop; however, the system flowrate is not affected because the pump has adequate margin to maintain the flow rate.

Therefore, the replacement valve will not increase the probability of occurrence nor the consequences of an accident evaluated in the accident analysis. The replacement valve will not

. change the operation of the system, will not introduce any new failure modes, nor change system operational parameters, Therefore, the replacement valve will not increase the possibility or consequences of an accident not previously evaluated. The replacement valve will not change RH performance criteria defined in the USAR or in the Technical Specifications, therefore the margin of safety will not be decreased.

Attachment 2 to U-602 36 Page 36 of142 i

NEW CONTROL VALVE (1CC080) IN COOLING LOOP RETURN LINE OF FC HEAT EXCHANGERS Document Evaluated: Modification CC021 Log Number: 96-023 This modification adds a new manual control valve to throttle the flow through the Fuel Pool Cooling and Cleanup (FC) System heat exchanger cooling loop. W manual valve will be used in i

lieu of the Component Cooling Water (CC) System Heat Exchanger Outlet Valves which are motor operated butterfly valves. The new valve will be installed in the CC return line riser, a non- ,

safety related portion of the CC system. Therefore, the addition of this valve will not increase the probability of, or the consequences of failures in equipment important to safety. Since the CC .

system is not required to maintain the plant in a safe condition in case of an accident, the modification will neither increase the probability, or the consequences of any accident previously evaluated. The new valve will not introduce any new failure modes and thus will not increase the possibility of any new accident or the consequences of any accident not previously evaluated.

Since no te,chnical specification is dependent on the non-safety related portion of *be CC system, the margin of safety will not be reduced.  :

REPLACEMENT OF THERMO LAG FIRE BARRIER IN CB 800' Document Evaluated: Modification FP-093 Log Number: %-025 R1 This modification replaces an existing Thermo-Lag fire barrier on conduits CO2999, IC03001, l and IC03002 with a 3-hour fire barrier made of concrete, mortar, and steel. The modification will eliminate all concerns about Thermo Lag in fire zone CB-6d, and will allow safe shutdown cables 1

- I AP28T, I AP341, and 1 AP28U to perform their safe shutdown functions in the event of a fire in zone CB-6d. The new fire barrier mee's the requirements of 10 CFR 50 Appendix R, section iii.U.2. The new Fire Barrier will act as seismic support SS F-1 for the conduits and their cables

,l AP28T, l AP341, and 1 AP28U.

The Safety Evaluation considered the construction activities as well as the post installation effects on instrumentation, control, ventilation circuits, as well as on control room integrity. The Safety Evaluation concluded that re-routing, cutting, or welding activities would not impact or affect any equipment important to safety which is powered or controlled by any cables, signal lines, or instrument sample lines. Therefore, the probability of failure or the consequences of failure of equipment important to safety would not increase. The Safety Evaluation also concluded that neither the construction activities nor final post installation fire barrier would affect or induce any spurious signal on any instmmentation or control circuits so as to act as initiator for any accident or its consequences previously evaluated in the accident analysis. Similarly, since operation of equipment served by any instrumentation, control, ventilation, or signal circuits is unaffected and l

no new failure modes or operating points are introduced by the construction or post installation modificction, the possibility of a new accident or its consequences is not increased. Since no operating parameters or setpoints are affected, the margin of safety is rot decreased.

= i_____________.___, __ _ _ _ _ _ _ _ . _

i Anaciuneet 2 i to U.<,02sw Page 37 of142 j

ELIMINATION OF STEAM CONDENSING MODE OF RH l t

Document Evaluated: Modification RH-033 Supplement.1 Log Number: 96-026 Modi 6 cation RH 033 removed the steam. condensing mode of Residual Heat Removal (RH).

This modification is issued in supplements that correspond to the phase of the modification that is l Implemented. Modification RH-033 Supplement 0 is reported by Safety Evaluation Log Number 93-115 earlier in this report. By letter dated April 10,1996, from D. V. Pickett, NRC, to M. W.

Lyon, Illinois Power, Clinton Power Station, the NRC iscued amendment 104 to CPS facility  ;

operating license NPF-62. Amendment 104 revised the technical specification in response to Modification RH-033. This supplement I to Modification RH-033 disconnects electrical and ,

control power to various motor operated valves (MOVs), eliminates computer points, makes l changes to the Safety Parameter Display System (SPDS), eliminates the Containment and Reactor ,

Vessel isolation Control System (CRVICS) signal fkom RH room HVAC to RCIC isolation, and adds a new vent to RH A to facilitate testing. The MOVs are effectively abandoned in place by

this supplement. The only change to the MOVs is the removal of power, signal, and control leads.- The MOVs will remain closed due to the self-locking stem / stem nut threads. The addition of the new vent to RH A constitutes addition of a small bore vent line and a small manual valve which is qualified to the same ASME code standards as the rest o.f the system. Amendment 104 to the Technical Specifications removed the requirement for the RH room HVAC to RCIC isolation from Technical Specification Table 3.3,6.1-1. This mo6ification supplement incorporates that change by removing the CRVICS isolation signa >

The effective abandonment in place of the existing MOV in a closed locked position and the removal of the CRVILS isolation signal authorized by amendment 104 to the technical Specifications will disable a mode of RH operation that has never been used, and as such will have no impact on the rest of the RH or Reactor Core Isolation Cooling (RCIC) systems. Similarly, the addition of a vent valve that is constructed to the same quality standards as the rest of the RH system will have no impact on the RH system. Therefore, Supplement I to Modification RH-033 will not increase the probability of failure of equipment important to safety, or its consequences.

USAR Section 15.2.4 discusses Main Steam Line isolation. The steam condensing mode of RH would' mitigate the sevuity of the transient by providing additional cooling. Without this mode of .

operation, the mitigation would be by cooling and depressurizing the primary system. The changes incorporated by this supplement would not act as initiator for the analysis of USAR ,

Section 15.2.4, and since alternate methods for providing this transient mitigation function are - ,

available, the probability or the consequences of an accident previously evaluated would not increase.= - Additionally, as the supplement does not add any new failure mechanisms, the possibility of a new accident, or its consequences, is not increased. Since the Technical

Specification Bases B 3.3.3.2 specifically exempts the RH steam condensing mode, the bases do not rely on this mode of operation for the RH system and the margin of safety is not decreased.

_e.m.,-.S. .. _ , . , , m.my y c.-,. n,... ,, .- . - ,, w y-c,',,,,,,,y.,-,..-3.m.., ,,,.nw..~m., r-.s m gmo r a*w--*=7

Attaciuneet 2 to U402sM Page 38 of142 REVISE USAR APPENDICES E AND F TO REFLECT THE NEW RADWASTE MACHINE SHOP CONFIGURATION Document Evaluated: Condition Report 196-05-016 Log Number: 96-027 This change revises the above listed USAR appendices to change fire zone R-li from low to

, moderate. The change and the re-claulfication of the fire zone are a result of additional i

equipment added to the radwaste machine shop. Additionally, fire protection Figures FP-18A &

B are revised for consistency with the parent station drawing.

Fire zone R li contains no equipment important to safety, and as such this reclassification of the fire zone will not increase the probability of failure, or the consequences, of equipment important to safety, The safe shutdown analysis remains valid and the probability of an accident, or its consequences, previously evaluated does not increase by this change. This re-classification of the fire zone does not introduce new failure modes or operating points and thus, the possibility of an accident, or its consequences, of an accident not previously evaluated does not increase by this change. Technical specifications are not dependent on the availability of this machine shop and the margin of safety is not reduced.

RPS REGULATING TRANSFORMER OUT-OF SERVICE FOR EXTENDED TIME Document Evaluated: Condition Report 1 M-06-032 Log Number: % 033 This safety evaluation reviewed the impact of operation with the regulating transformer out of service for an extended time. The transformer is out-of service due to maintenance. The self-regulating transformer provides an alternate source of power to the non-divisional Reactor Protection System (RPS) inveners. Without an alternate source, the non-divisional power supply inverters would be de-energized if the preferred source were unavailable. These inverters are required in MODES 1, 2, and 3 'and the station enters LCO 3.8.7 C if unavailable in those modes.

Unavailability of the transformer would increase the likelihood of an unavailable inverter but does

- not increase the probability of failure of the inverter. Unavailability of the transformer will not impact the control and protective functions of the RFS, as the transformer provides the alternate source. Losses of one supply to the inverters will neither cause, nor inhibit, a protective action.

Thus, the probability of occurrence or the consequences of an accident previously evaluated will not increase. Unavailability of the transformer does not introduce any new failure mode or operating point, thus the possibility of an accident, or its consequences, not presiously evaluated is not increased.

Operability evaluation condition report (CR) 1-96-06-032 determined that the inverter should be considered operable even though the transformer, which supplies attendant power, is unavailable.

The requirements of LCO 3.8.7.C are met and the margin of safety is not decreased.

Attachen4 2 to U 602836

, Page 39 of142 REACTOR COOLANT SYSTEM CHEMISTRY Document Evaluated: ORM Section 2.3.1 Log Number: 4 034 This change to the Operational Requirements Manual (ORM), removes routine pH determinations (at least once in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) but maintains the sampling requirement for pH when the reactor coolant conductivity is above the limits of the ORM Table 3.3.1-1. This change eliminates a redundant water chemistry measurement. Because of the difHculty of accurate pH measurement in pure water, the BWR Water Chemistry Guidelines do not require routine collection ar.3 reporting of pH data and only recommends sampling and pH measurement as a means of verifying the presence of abnormal chemistry conditions and diagnosing the cause. Conductivity provides a good and prompt measure of reactor water quality. Reactor water conductivity in its normal range is an indication that pH, chlorine, and other impurities affecting the conductivity measurement are also within ti..ir normal range. Significant changes in conductivity readings provide an indication that fuither water analysis is required. The iate of hydrogen generation following a LOCA is influenced by the pH of the primary coolant. The operational limits placed on pH are based on the assumptions used when calculating corrosion rates and thus hydrogen generation rates. This change is consistent with the EPRI BWR Water Chemistry Guidelines.

Removal of the routine pH analysis requirements when conductivity is within the limits specified in the ORM, and there is assurance that the coolant pH is within the limits ORM Table 3.3.1-1, will not result in an increase the rate of metal corrosion. Thus, removal of the 72-hour sampling requirement will not increase the probability of failure or consequences of equipment important to safety. USAR Section 6.2.5 evaluates the effect of hydrogen buildup on the containment and daywell following a LOCA. Removal of the 72-hour sampling requirement will not impact the probability of a LOCA, or lead to an increase of the hydrogen generation rate, wMeh would increase the consequences of a LOCA. Sampling pH only when conductivity iwutside the specifications of ORM Table 3.3.1 1 will not increase the hydrogen gener:'. tion rate and increase the combustible gas concentration in containment or the drywell, thus the possibility of an new accident or its consequences will not be increased. Using conductivity as an indicator to determine if pH is within specification will not result in an unnoticed 'out of specification pH, and the assumptions of the post LOCA hydrogen buildup analysis are maintained, thus the margin of safety is not reduced.

DELETION OF AREA RADIATION MONITORS IN THE RADIOLOGICAL CONTROL PROGRAM Document Evaluated: USAR Change Package 7-119 Log Number: % 036 The USAR change revises Section 1.8 to take exception to Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Exposure at Nuclear Power Stations Will Be As Low As Reasonably Achievable," position C.2.g, reference 10. Clinton Power Stat ion is committed to Revision 3 of the Reg. Guide 8.8. The regulatory piiide references ANS/HPS 56.8 " Location and

- Design Criteria for Area Radiation Monitoring Systems for LWR." ANS/HPS 56.8 was used by CPS to determine initial placement of Area Radiation Monitors. This ANS!HPS standard was never issued. However, CPS has adopted ANSI /ANS-HPSSC 6.8.1-1981 " Location and Design

Acachnunt 2

so vso2su ,

Page 40 of142 j

, . Criteria for Area Radiation Monitoring Systems for Light Water Nuclear Reactors" as the _ i j guidance for placement of area radiation monitors. In addition to the USAR changes, several area l I

radiation monitors are abandoned in place. The abandonment of the area radiation monitors was j justified either by: 1) the criteria of ANSI 6.8, Reg. Guide 1.97, " Instrumentation for Light.  ;

j Water-Cooled Nuclear Power Pla.nts to Assess Plant Conditions During and Following An  ;

i Accident;" 2) changes in station conditions such as decommissioning of equipment; 3) change in i

! use of fhcilities and areas; 4) historical surveys, or; 5) the availability of duplicate monitoring t i instruments. The purpose of area radiation monitors is to warn personnel ofimmediate changes in

! radiatior Wis. The areas where the monitors were abandoned were all evaluated for activities, j which woud cause immediate changes in radiation levels. The evaluations concluded that the >

abandonment of these monitors did not pose a personnel hazard and would not increase the -

F potential to violate the limits of 10CFR 20. '

i i

SPENT FUEL COOLING CHANGE TO ALLOW BOTH PUMPS TO BE REMOVED FROM  !

SERVICE DURING MAINTENANCE i l Document Evaluated: USAR Section 9.1.3.4 Log Number: % 037 f

l- The change revised the USAR description of the Fuel Pool Cooling and Cleanup (FC) System.  ;

i- The system is required to remove decay heat, maintain water level, clean up the pool water to .

! minimize fission product concentration, and maintain water clarity. Some valves are common to j both trains and both pumps must be secured to perform maintenance of these valves. This change j allows both trains of FC to be removed ikom service for maintenance as long as spent fuel pool i water temperature does not exceed 120'F.

j The compt ents important to safety in the FC system includes containment isolation valves, i valves tha maintain fuel pool level,- and the valves, pumps, and heat exchangers used to maintain '

fuel pool temp-ratures. The short duration maintenance outages where the pumps from both

! trains are secured will not impact the equipment required for containment isolation or the level -

l control function. Securing the pumps from both trains for a short duration maintenance outages -

g where the pool temperature is not allowed to increase above 120*F does not impact the

! _ equipment required for containment isolation or for pool level control. Therefore, short duration ,

i outages for performing maintenance on valves common to both trains will not increase the probability of failure of equipment important to safety. As temperature and level are monitored t

during these maintenance outages, the consequences of failures will not increase. Securing pumps firom both trains for a maintenance outage will not act as an accident initiator, iflevel or temperature have to be controlled during the outage,' the Shutdown Service Water (SX) System is

! used in a feed and bleed type mode. Therefore, securing pumps rtom both trains will not increase

[ the probability or the consequences of accidents previously evaluated._ Securing both FC pumps i-at the t.ame time stops forced circulation._ USAR Section 3.1, Attachment A3,1, Section 3.8

summarizes detailed thermal hydraulic calculations that show that with natural circulation and j 150*F water no localized boiling would occur at the hottest fuel assembly. Therefore, securing both FC pumps would not increase the possibility or the consequences of a new type of accident.

I Since temperature and level in the spent fuel pools will remain within their normal ranges during

! short-duration maintenance outages, the margin of safety will not be reduced. {

N 5

.a_ _ _ , . _ _ . _ _ . . _ _ . , _ . - _ . , . . . _ _ . . _ _. _ - , . ___.______._____._.__.;__._-.. _ , . , . . .

Anh 2 to U4028M -

Page 41 of142 ISOLATION VALVE

SUMMARY

FOR LINES PENETRATING CONTAINMENT Document Evaluated: USAR Table 6.2-47, and ORM Attachment 4 Log Number: 96-038 This safety evaluation supplements Safety Evaluation (SE) 95 078 by showing another instrument

-line through penetration IMC 169 in the USAR table. This change does not invalidate the analysis of SE 95-078, but it changes the calculation that was performed in support ofit showing that the calculated leakage was below the Technical Specificatiorilimit of 908 A' / day. The safety evaluation was performed because SE 95 078 assumed that only IMC-169001 was a %

- inch line and the only one in the penetration. The penetration contains two % inch instrument lines. The calculation in this SE accounts for the removal of IMC-079 and for two % inch lines through IMC-169 instead of one % inch line. The calculation showed an effective total potential bypass leakage path of 811.2 A' / day, an increase of 0.6 A8 / day over the calculation of 95 078.

This USAR revision, to include the second instrument line in penetration IMC 169 is administrative, and does not increase the probability of an accident or its consequences which was previously evaluated in the USAR, nor is the probability of a failure of equipment important to safety increased. Since the additional leakage path attributed to the second instrument line

- produced an effective leakage ofless than the Technical Specification limit, the margin of safety is not reduced.

DELETE SILICA MEASUREMENT REQUIREMENTS FROM SPENT FUEL POOL COOLING SYSTEM DESCRIPTION Document Evaluated: USAR Section 9.1.3.2 Log Number: 96-039

. This change removed the requirement for determination of silica on the effluent of the in service

[

fir? pool domineralizers Rom USAR Section 9.1.3.2. Limits for conductivity, pH, chloride, and silica were listed in USAR Section 9.1.3.2 for demineralizer effluents. These limits, according io j: - the USAR, are to ensure the maintenance of the quality of water in the fuel pools, by indicating

' demineralizer efficiency. The demineralizer resins are replaced when conductivity, pli, chloride,

or silica is outside of the limits of USAR Section 9.1.3.2. EPRI BWR water chemistry Guidelines

. state that silica removal is not via an ion exchange mechanism. Thus, efficiency of the domineralizers cannot be determined by the silica concentration in the effluent. Further, the USAR specifically exempts silica from the list ofitems that the fuel manufacturer recommends should be used to ensure the quality of water in the fuel pools. Removal of this diagnostic parameter kom fuel pool domineralizer efficiency does not adversely aff'ect the water chemistry or the quality of the water in the fuel pools. Therefore, the removal of silica as a measured variable L does not increase the consequence of any accident, nor does it increase the probability of l accidents evaluated or possibility of a new type of accident. Similarly, the deletion of silica as a i measured variable will not decrease the margin of safety.-

1- ,

4 t

~

Anachment 2 to U402s%

Page 42 of142 REMOVE WS NaOCl INJECTION PUMP FROM WS SYSTEM Document Evaluated: ECN 29774 Log Number: 96-041 This ECN removes the Plant Service Water (WS) System NaOCl injection pump (1WS06P) from the Circulating Water Screen House (CWSH). This pump had been installed by modification WS.

021, reported in USAR Revision 5, SE Log Number 93-0023. This pump was added in the spring of 1993 as an interim method of chlorinating WS without discharging to the Ultimate Heat Sink (UHS). The backup method of chlorination will be unavailable with this pump removed.

However, the pump is no longer necessary due to modiAcation CL-007 (reported in this USAR Revision, Log No.95-070). This pump has not been used since the fall of 1995. This change will only affect USAR Figure 9.2-1, Sheet 1. This change does not impact the description of the WS system in USAR Section 9.2.1.1. The WS system is not required to ensure safe shutdown of the plant and is not designed to meet seismic Category I events. NUREO 0853," Safety Evaluation Report Related to the Operation of Clinton Power Station, Unit No.1," identifies WS as a nonessential plant service water system. NUREG-0853 only specifies WS is able to separate from the Shutdown Service Water (SX) System and will have procer udiation monitors that can bolate WS on a high-radiation alarm. This change will not affect these fbnctions. As a nonessential system, this change would not impact the margin of safety.

FEEDWATER VALVE REPLACEMENT (IFWO21)

Document Evaluated: Modification FW 036 Log Number: 96 042 Modification FW-036 replaces the 16-inch nominal ploe size long cycle flush valve IFW-021.

The old valve la a motor operated Y pattern globe valve. The new valve is a motor operated angle valve with a CCI DRAG' disk stack. IFWO41; a four-inch globe valve, will also be removed. The new IFW-021 valve will allow throttling through the full range ofilows of the old pair of valves. The restriction orifices will be removed and the IFWOO4 valve will not need to be throttled. The change allows restoration of original design basis long cycle flushing flows of kpproximately 8,000 gpm without vibration. The modification includes pire hanger changes. The power supply for the IFWO21 valve will be changed. The Main Control Room (MCR) benchboard will have the position light for the IFW016 valve removed and the handswitch for the old IFWO21 valve will control the IFW016 valve. The handswitch for the removed IFWO41 valve will control the new IFWO21 valve. This change will revise USAR Figures 3.6-1, Sheet 14; 10.1 1 and 10.4-7.

The change affects non-safety related high energy pipina. The design requirements (seismic, structural, ASME) for this will remain as the original design. Therefore, the probability of a high energy line break would not change. This design simplifies Feedwater (FW) System operation, Therefore, probabi ity of operator error would be reduced. The design basis event in the USAR Chapter 15 is a 20-inch FW line break close to the safety related portion of FW piping in the steam tunnel. Failure of this 16-inch line would still be enveloped by the 20-inch line break.

- Therefore, this change would not impact accidents evaluated as design basis. The only potential accident of this non-safety portion of the FW system is a line break, which has already been 4

l AM=agt 2 to U4028M Page 43 of142 coasidered in the design basis. Therefore, there is no potential for a new accident type. Since the change is bounded by other more limiting events, there is no decrease in the margin of safety.

CHANGE MINIMUM ALLOWABLE SPENT FUEL POOL TEMPERATURE Document Evaluated: USAR Section 9.1.2.3.1.1, Attachment A3.1 Log Number: 96 044 1

This change alters the bases of the design analyses, summarized in USAR Attachment A3.1, for J the spent fuel pool. 'Ihis lowers the minimum operating temperature firom 68 'F to 40 'F. This change la needed since the normal Component Cooling Water (CC) System is unavailable during a Plant Service Water (WS) System outage. When this occurs, the Shutdown Service Wa;er (SX)

System supplies the cooling water for the spent fbel pool heat exchangers. The SX temperature will be below 68 'F when a WS outage would be planned durin6 a refbeling outage. Attachment ,

~

A3.1 spent fuel pool storage rack design information is being moved either into the USAR within text orin the reference section. i This ciense was evaluated by General Electric (GE) for effects to piping, filtration, monitoring instrumentation, area radiation monitor and k-effective impacts. The systems and system interactions, thermal, material and mechanical, were not affected by changing the temperature to 40 'F. As such, there was no increase in probability of failures associated with thl change. The i USAR Chapter 15 Fuel Handling Accident would not be affected by a change in water temperature. The Bases to the CPS Technical Specification has not been impacted by this change.

Therefore, although the k-effective of the loaded spent fbel racks would be slightly closer to the 0.95 limit, the :nargin would not be reduced.

GENERATOR STATOR WATER COOLING AIR INJECTION TEST -

Document Evaluated: CPS Procedure 2800.71 Log Number: % -045 ECN 29644 will install 3/8 inch copper tubing, valves and a flow meter to iriject instrument air L

into the main genera;or stator cooling water. This will be implemented in order to increase the dissolved oxygen content to the original equipment manufacturer recommended value of two to eight ppm. CPS Procedure 2800.71," Generator Stator Water Cooling Air Injection Test,"is a new special test to allow injection of air at certain pre-determined levels and records dissolved oxygen level in the stator water. After completion of testing, the results w*i. 0c used to revise L CPS Procedure 3110.01," Generator Stator Water Cooling (GC)." This special test procedure allows irdection of air in steps of about 200 ml/ min. and records the change in dissolved oxygen levels.; The change in chemistry will take several days to respond due to the small quantity of air injected. The conductivity of the stator water will be continuously monitored during the test. If these chemistry changes cause problems, the situation would be seen and transients avoided. As such, the generator would not be in risk of new tratulents involved with this test. The existing protection of the generator by interlocks,and trips is adequate. The GC system la a non-safety ,

I system. The USAR does not have equipment malfunctions evaluated for the GC system. The small quantity of air being injected will be absoited by the stator water withmt producing new I '

- accident types or malfunctions.

e

, . . - . .v__,,,,...i,. ., ,..,._,m_,,,_,_m_ ,m_____,,___,._m_.. . ~ m__y .__,___,__.._,.,-m..,

Attachinent 2 toU4028%

, Page 44 of142 4 CORRECYlON OF USAR DISCREPANCIES REGARDING FUEL RECEIPT ACTIVITIES Document Evaluated: USAR Section 9.1.4 LogNumber: 4046 This changes fi'.al!sJung niethods described in the USAR. US AR Section 9.1.4.2.10.1," Arrival of Fv4 cA Slic/* has a statement that the outer containers are left on the trailer. This sentence

.was deleted as there is insufficient room to safely handle wooden shipping crates on the trailer.

USAR Section 9.1.4.2.10.2.1.1, " Receipt and Inspection of New Fuel," was revised to reflect the fhet that new fbel is placed on the fbel handling floor using the auxiliary hoist on the fuel building crane. In addition, the first paragraph of this section was deleted and replaced with 9.1.4.2.10.1 information which more accurately describes CPS. This section's last paragraph which discusses new fuel receipt, storage and inspection is being deleted. This is being done since the method described is not as practiced at CPS and is redundant to the accurate description of this activity given in USAR Section 9.1.4.2.10.1. USAR Section 9.1.4.2.10.2.1.2, " Channeling New Fuel,"

has been changed to consolidate and clarify discussions regar6ng the installation of new and used -

(irradiated) channels. Wording has been added to better identify the two processes. Wording -

from USAR Section 9.1.4.2.10.1 was moved to this ocction to consolidate $1milar discussions.

These changes would not impact probability or consequence of equipment malfunctions as

- equipment is being used in a manner consistent with the USAR description. The bounding accident is the impact of a channeled bundle onto an un-channeled fuel bundle in the spent fuel pool. This is unchanged as irradiated fuel is not being handled during new fuel receipt. The margin of safety is not impacted as these changes can only aff'oct new (non Irradiated) fuel. >

ALTERNATE SHUTDOWN COOLING WITH IRRADIATED FUEL IN UPPER POOLS Document Evaluated: CPS Procedure 3312.02 LogNwnber: %047 Procedure Advance Change (PAC) 0432 96 revised CPS procedure 3312.02," Alternate Shutdown Cooling (A SDC) Methods," to allow an attemate shutdown cooling method with irradiated fuel in the temporary fuel storage racks in the containment steam dryer storage pool.

- This method consists of using an Emergency Core Cooling System (ECCS) pump, one train of Suppreanlon Pool Makeup (SM) upper pool dump valves and the suppression pool cooling 3 system. The change requires that the gate between the steam dryer pool and the reactor refueling pool be removed prior to use. The procedure will indicate that the use of this method will add approximately three feet of water to the suppression pool level. However, the procedure requires

greater than three feet of freeboard (wall between pool and drywell) before the procedure can be implemented. . This charge was reported in USAR Revision 4 (Log Number 92-0031) except for allowing spent fuel to be in the upper pool.

An inadvertent upper pool dump with fuel in the upper pools has already been evaluated in USAR Section 6.2.7.3.3. This procedure (step 8.4,2.c) already invokes the ORM requirement 2.4.8 that i .this mode of operation cannot be initiated until the irradiated fuel has been secured in the temporary storage racks or in the reactor pressure vessel (RPV). With irradiated fuel handling suspended, adequate shielding (approximately twenty feet of water) will cover the irradiated fuel.

__..__._m..__.-___ __ __._ _ _.___ _ _ _ _ _ _ _

Aten.e 2 to U4028%

Pces 45 of142 Therefore, fuel barrier cladding analysis would not be affected and margin of safety would be j

unchanged. Since this chuge ofJy aDown spent fuel in the upper fool utilizing this A SDC 1 me4m, and thic is already analyted in the USAR, this would not create a condition outside that alteey waleated !n the USAR. %erefore, accident types or frequencies of accidents would not be Chang (M. <

DE RATlhO Ott 13 M#uf TON HOISTS TO NUREG fWin SINGLE FAILURE PROOF -

EQUIVALENTS  ;

Document Evaluated: USAR Section 9.1.4 Los Number: 96 048 I This Safety Evaluation (SE) evaluates a USAR revision that de rates the Fuel Building crane 125

- ton main hoisi to 62 ton 6 and the 10 ton auxiliary hoist to 3.27 tons. This SE also evaluates a propond revkica to the USAR that de-rates the Containment Building Polar crane 10 ton y' auxiliary hoist to 3.27 tons. These changes are to align the USAR to commitments made to NRC Bulletin 96-02, " Movement of F 4 wy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety Related Equipment." This will also comply with NUREG 0612 " Control of Heavy Loads at Nuclear Power Plants, Resolution of Generic Technical Activity A-6." During CPS l initial licensing, it was identified that then hoists were not designed or constructed to be fully L single failure proof, however, the USAR was not revised. As a result of an 19891990 Independent Safety Engineming Group (ISEG) evaluation the holsts were administratively de-

rated. Tids change would not irrpset equipment malfunctions since decreasing the rating capacity 4

would lessen the likelihood of equipment failures. Additional or new accidenta would not be f expected as the equ!pn ont is more able to carry the de-rated load. Margin would not be decreased because a hoist carrying less of a load is less likely to fail and damage protective barriers.

. REVISION TO RCA ACCESS TRAINING REQUIREMENTS Document Evaluated: USAR Sections 13.2.1 and 13.2.2 Log Nu:nber: %-050 These changes add a new training course that allows noa radiation wonkers, not performing ,

radiological work, access to the Radiologically Con + rolled Areas (RCA). This la in accordance with 10CFR19.12 and does not conflict with CPS commitments to Regulatory Guides 8.27, "Ra6ation Protection Training for Personnel at Light Water-Cooled Nuclear Power Plants," and 8.29," Instructions Concerdng Risks From Occupatloal Radiation Exposura" The change also

' removes the annual requirement for refresher training of radiation work practical demonstration -

training. Personnel will complete refresher training based on a self usessment of dressout skills.

A statement has been added to provide radiation worker training for temporary personnel

- commeneurate with the level of radiation work to be performed. The description of abbreviated Respiratory Protection Training was also changed to be conslatent with other weess training programs. This would not impact equipment malfunction types or frequencies. This change does l not effect the training for maintenance ofimportant to safety equipment. Consequences of i accidents would not be affected as personnel with unescorted access to the RCA are trained in proper response to plant alarms. These access training changes would not affect protective

. barriers or margins.

L

.____u,_ _ -.____,;,_.-.________.-___.,,....____,- - - ~ .

W ,

Attachment 2 i to U 6028M  ;

Page 46 of142 j l

3 ,

- PLACE NINE RA SYSTEM ANNUNCIATORS IN THE MCR "NOT IN SERVICB" 4

Document Evaluated

ECN 29845 LogNumber: 5051

This change places nine Breathing Air (RA) System auunciators in the Main Control Room i (MCR)"Not in Service." This is being done because the alarming portion of the RA system is no l

, longer being used and the annunciators are a distraction to operators. The supporting drawings,  !

4 procedures, USAR and training information is being updated to reflect this change. This will '

{ require revising USAR Section 9.3.1.5 and Figure 9.3 3, Sheets 2,3,4,5 and 6. h deletion of these nine annunciators will not increase the probability or consequences of a malfunction of i equipment important to safety. With the exception of the MCR portion of this system, the l 4

remainder of the RA system has no nuclear safety related functions. All of the system, except the MCR portion is not used and is planned to be abandoned. USAR Section 9.3.1.3 states that failure of the RA system will not compromise any nuclear safety-related system or prevent safe -

f shutdown. Therefore, new failure modes would not be applicable and margin of safety i - unaffected.

_ IPA 08J POWF,R SUPPLY PS2 TEMPORARY MODIFICATION  !

i r i

Document Evaluated: TM E055 LogNumber: %-052 L i To facilitate replacement of the PS2 power supply in cabinet IPA 08J, the Turbhe Building Closed Cooling Water (WT), Component Cooling Water (CC) and Make Up Water Pump House level transmitter leads are required to be lifted and a jumper on a Cycled Condensate (CY) relay L installed. Performing this activity operates equipment dtTerently than described in the USAR for

- the WT and CC systems, Level transmitters for these are described in USAR Sections 9.2.2 and E 9.2.7 as having a trip function. N WT transmitter trips the pump on a " low low" tank level. ,

a h CC transmitter trips the pump on a " low" tank level. This TM defeats the trip to the pumps  ;

as described in the USAR. All four systems affected by this TM are not important to safety. As

, such, this would not impact the accidents, malfunctions or cause new failure modes for safety

' related equipment. The CPS Technical Specifications and Bases do not contain this level of detail i

regarding this TM scope. Therefore, the margin of safety would not be negatively impacted by the TM.

CHANGE THE USAR DESCRIPTION OF NUMBER OF NON LICENSED OPERATORS NEEDED 4

I Document Evaluated: USAR Section 13.1.2.3 Log Number: 96-053 R1 This change revises CPS checklist 3800.02C002, "D-Area Daily Rounds Sheet." This is an administrative log taking procedure used to verify the proper operation of systems. This change comblaes the log taking activities of two operators into one. The change also removes the Radwaste Operator position from the USAR. These duties are being performed by the D-Area

, Equipment Operator. The D-Area Equipment Operator is capable of performing the D-Area i tours and the tours of radwaste equipment. h practice of performing rounds is a passive 2

I 1;

,. __,-.4, - - . . - _ , , _ _ - . - . _ . . . . _ _ _ _ _ _ . .m.- --,._.__ _ _ - . . . _ , _ - . ~ . . - - . - - - - .

Attaciuneet 2 to U402:36 Page 47 of142 activity which would not increase the probability of equipment failures or malfunctions or equipment malfunction consequences. Since equipment failure rates are unaffected, the probability and consequences of an accHent would not be changed by combining these duties.

LONG TERM TAGOUT OF OFFOAS (00) GAS COOLER DEFROST HEATER Document Evaluated: Tagout #951142 Log Number: 96 056 This evaluation is for maintaining the 00 gas cooler defrost heater (IN66 Boll) danger tagged out of service for an extended period of time. As corrective action for CR 1 92-05 0$9, CPS procedure 3407.01, "OG Vault Re&igeration System," is being revised to prevent operation of the heater. The gas cooler is described in USAR Table 11.3 2. The use of a blast heater will accomplish the 00 gas cooler de&ost heater's fbnction.- USAR Section 9.4.8.2 states, "as it may occasionally be necessary to heat the vault to 150'F by using an electric-blast coil and a heating-circulating fan laterconnected to the re&igeration system ductwork." As such, this actMty is already evaluated within the USAR. The most severe accident that could happen is a loss of  !

condenser vacuum if the 00 cooler freezes and prevents off' g as flow through it. This has already l been evaluated in USAR Section 15.2.5. The probability of this event will not be increased.

TEMPORARY MODIFICATION OF THE AR/PR SYSTEM CENTRAL CONTROL  :

TERMINALS Document Evaluated: TM 96-070 Log Number: 96-057 This Temporary Modification (TM) removes the monitor and disk drive unit kom the Padiation ,

Protection (RP) office Central Control Terminal (CCT), removes one disk drive nom the Main Control Room (MCR) CCT and restores the capability to transfer the Area Radiation / Process )

' Radiation (AR/PR) system monitoring function to the MCR CCT, In addition, this TM restores -

the capability of the Technical Support Center (TSC) Cathode Ray Tube (CRT) to operate l utilizing the required components from the RP office and MCR CCTs. The TM also modifies the data source circuitry to the Nuclear Data computers to include a second switch which will permit the selection of either the MCR CCT or the RP office CCT as the primary source of APJPR data to the "outside" organization as required. Afier this TM is implemented, the system configuration will consist of: An operable CCT in the RP office which will be in service, capable of providing AR/PR information; A CCT in the MCR capable of being placed into operation if required; A ,

CRT display in the TSC capable of being placed into operation if the MCR CCT is placed in operation. This TM is required to comply with systera requirements until modification PR-F018 is implemented.

This change will not aff'ect the reliability of the CCTs;'they will continue to be operated in the ,

same manner. As such, this TM would not increase the probability of malfunctions of the CCT.

The accidents analyzed in USAR Chapter 15 do not depend on operator action based on AR/PR system s,larm annunciation. Therefore, this change would not result in an increase in the probability or consequences of an accident previously evaluated in the USAR. This change will not eff'ect the reliabiiity of the AR/PR system thereby not producing new accidents not already evaluated in the USAR; The information provided by this system and the CCTs is not used by the 1

-- - , . _ _ _ _ _ . _ __ _~ - _ _ , ~ . - - , - _ _ . _ _ , - - _ , - - . . _ ,

Anachment 2 to U 602836 Page 48 of142 operators to determine proximity to any CPS safety limits. Therefore, margin of safety is not impacted by this TM. This analysis is elso valid for safety evaluation (SE)96-060.

REMOVAL OF STEAM BLANKETING FROM CPS PROCEDURES 3002.01,3006.01 AND 3105.01 Document Evaluated: CPS Procedures 3002.01,3006.01,3005.01 LogNumber: 96-058 This removes references to steam blanketing system in CPS procedures 3002.01, "Heatup and Pressurization," 3006.01, " Unit Shutdown," and 3105.01, " Turbine (TO, EHC, TS)." USAR Sections 1.2.2.8.13 and 9.5.9.2 state that steam produced by the auxiliary steam boilers are used during startup and shutdown to provide blanketing steam to the Moisture Separator Reheaters (MSRs), Additionally, USAR Table 9.2-1 lists the steam blanketing drain cooler. W future use of steam blanketing system will be decided based on low pressure (LP) turbine inspection in RF-6 and the decision to replace LP rotors which is tentatively planned for the eight refueling outage (RF-8). The steam blanketing system is designed to prevent rust from forming in the MSRs. The steam blanketing system is presently non Ametional. This safety evaluation is to address removal of steam blanketing.

As defined in USAR Section 9.5.9.3, the auxiliary steam system has no nuclear safety related components. Failure of this system would not impact safe shutdown of the reactor nor increase the probability of malfunction or consequences of accidents with safety-related systems. Also, this would not create the possibility of new accident types for safety-, lated equipment. As stated in 9.5.9.3, this syste n does not have safety-related components and would thereby not affect fbel barrier or the margin of safety.

INADEQUATE SAFETY EVALUATION SCREENINGS FOR TEMPORARY MODIFICATIONS Document Evaluated: CR 196-10-007 Log Number: 96-059 While performing the review of the safety evaluation screening for Temporary Modification (TM)96-059, Revision 1, it was discovered that Figure 8.3-1 of the USAR was affected by this TM.

CR 1 96-10-007 was written to address the fact that a safety evaluation was not written for this TM and TMs95-057,95-058,95-060,95-061 and 95-%2 without adequately addressing that the confi;~ ration of Figure 8.3-1 in the USAR was impacted by the TMs. USAR Figure 8.3-1 is the electrical loading one-line diagram that shows all of the major buses and loads at CPS. Due to the bus outages in RF-6, TMs listed above were written to supply temporary power to loads that were needed for the outage. The normal supplies to these loads were all non-safety and the temporary feeds were also non safety. In addition, all the equipment is normally fed from the non-safety 6.9 KV bus l A. Therefore, these TMs would not affect consequences of failures or malfunctions of safety-related equipment in the USAR. The temporary power supplies operated in the same manner as the normal sources; as such, probability of failures would not be affected.

Seismic and st'ructural interactions were also evaluated acceptable. Therefore, this is no potential for creating new accident types. Since these TMs were only affecting non safety equipment, the margin of safety would not be affected.

Attachment 2 to U402:36 Page 49 of 142 REMOVAL OF RP OFFICE CCT FROM SERVICE i

Document Evaluated: USAR/ODCM Log Number: 96-060 i i

Temporary Modification (TM) 5070 (reported earlier) changed the configuration of the Area l

Radiation / Process Radiation (AR/PR) System and associated Central Control Terminals (CCT). i Specifically, the CCTs, loceed in tne andlation Protection (RP) omce and the Main Control  !

- Room (MCR), do not contain a CRT display, it was determined that two CCTs would not l function properly with both operabb at the same time. The CCT in service for some time has j been the RP office CCT. It was decided that the RP office CCT would be the operable terminal ,

since there wu a concern regarding the number of" nuisance" alarms nom the system in the  !

MCR. Recently, it has been determined that the MCR CC r should be the only operable terminal.  ;

Therefore, to address the number of alarms in the MCR it has been decided to reduce the - l population of mon l tors polled by the CCTs. The USAR and Offsite Dose Calculation Manual  !

(ODCM) currently indicate that there are redundant CCTs and that information from the AR/PR

system can be obtained hom either CCT as needed. The USAR also indicates that the CCTs are i equipped with a CRT display and that certain monitors are polled by the CCTs. It is therefore l necessary to revise the USAR and ODCM to reflect the change in system configuration This  :

i safety evaluation (SE) is for the USAR changes. The safety significance of any physical changes  !

to the plant was evaluated in SE Log Number %057 for TM %070.  ;

l RADIATION PROTECTION OFFICE CENTRAL CONTROL TERMINAL (CCT) REMOVED l FROM SERVICE -

i  ;

i Document Evaluated: Emergency Plan Advance Change Notice Log Number: 4061 ,

I (ACN) I!!!2 i This ACN revises Emergency Plan Sections 3.2.5 and Table 1.5 because of recent plant

! . modifications, This ACN deletes reference to the Radiation Protection (RP) Office CCT hom

Section 3.2.5 of the Emergency Plan and revises the description of the source of effluent monitor ,

i ~ information to the Safety Parameter Display System (SPDS). This deletion of the reference of the RP office CCT from the Emergency Plan and the change in description of the source of effluent information to the SPDS are in support of Temporary Modification 96-070 (Los Nu'nber

%057). Temporary Modification %070 restored the Main Control Room CCT and the Technical Support Center (TSC) Cathode Ray Tube (C.RT) to service. The Temporary

. Modification moved computer peripherals from the RP Office CCT to the Main Control Room CCT to return the Main Control Room CCT to service. The Temporary Modification also configured the CRT in the Technical Support Center (TSC) as a remote terminal. Changes to the USAR and the ORM were evaluated under SE Log Number 5060, also reported earlier in this

! report. Additionally this ACN revises Table 3-5 of the Emergency Plan for consistency with

, USAR Table 12.3-2. The ACN removes reference to several Area / Process Radiation Monitors.

I The removal of the radiation monitors from the USAR Table was evaluated'by Safety Evaluation

- Log Number 5036, also reported earlier in this report, The safety significance of any physical changes to the plant and changes to the USAR and ORM were evaluated by the above listed safety evaluations.- As such, this change is an administrative change which revises the description 4

a em,-men m -w--,.,w- e-,, .e --wr>. , +,w -

r-r w _ . w , r r o w , --* ev-

1 l

Anachment 2 to U402436 Pass $0 of142 of the CCTs available during activation of the Emergency Plan (Section 3.2.5) and the description j of the location of the radiation monitors (Table 3.5.) i The changes provided by this ACN do not provide instruction for equipment operation or l introduce any type of fhilure mechanism into the operation of safety related equipment. ,

Therefore, this ACN does not afect the probability of failure or the consequences of failure of equipment important to safety. The Emergency Plan is used to respond to abnonnal plant i conditions. This ACN does not affect the response or the ability to mitigate those conditions. . i The change to the Emergency Plan does not impact the consequences of any previously evaluated l accidents or of any new accidents. As this ACN is solely administrative it has no impact on the .

probability or possibility of any accidents new ot previously evaluated. The margin of safety is  !

similarly unaffected by this ACN since it has no impact on the basis for any technical -

specifications.  !

REACTOR RECIRCULATION (RR) PUMP MOTOR LUBRICATING OIL CONSTANT  !

LEVEL OILER DEVICE l Document Evaluated: ECN 29742 Log Number: 96-062 The change adds a constant level oiler device to each RR pump motor lower bearing lubricating oil system. The oiler device was added as a proactive measure to make up for minor oil consumption on the RR pump motors. The pumps have experienced low oil alarms. Routine j; inspections performed during RF-4 and RF-5 revealed a fine film of oil in the area around the i lower bearing oil reservoirs.

i USAR Section 15.3.3 identines fhilures internal to the RR pump, whereby the pump or motor

seizes. This mode offailure is not applicable to this change. The oiler device is an additioral oil ,

l reservoir for the lower motor bearing. Gross failure of this additional reservoir would not deprive

( the lower bearing oflubrication as the oil level would only drain to % inch above the oil supply j hole to the bearings and would not cause RR pump failure. USAR Sections 15.2 and 15.4 also identif'y failure of the flow control valves in the RR system and a pump shaft break. These failures would not occur as a result of a gross failure of the continuous oiler system. Therefore, the 2 addition of the oiler device would not increase the probability of, or the consequences of, faliure of equipment important to safety. USAR Section 15.3 analyses a decrease in reactor coolant i.. system flow rate. . Under this accident, a decrease in coolant flow rate due to trip of a RR pump can occur due to internal failure of the pump or due to motor overload or over current protection.

Since failure of the oiler will not cause the bearing to lose lubrication these causes o. U, pump j trip will not result from a gross failure of the oiler. Additionally, USAR Appendix E hection 3.3.1 ,

f discusses a fire in the drywell and identifies the combustible materials in this area. The j temperature characteristics of the material the oiler is made of, the oil flash point, the normal

, temperatures in the daywell preclude the continuous oiler from initiating a fire in the drywell and from failure due to high operating temperatures. Therefore, the oiler does not increase the 4- probability or the consequences of previously evaluated accidents. The oiler design and construction meets the same seismic requirements as the RR pump. The additional lubricating oil as combustible material was accounted for in a station calculation. A resiew of the materials of d- construction was performed to determine if any material incompatibilities existed. Since the

Attachment 2 to U402836 Page $1 of 142 materials of construction are not incompatible, the design and construction is the same as for the RR pump, and the additional lubricating oil as combustible material does not invalidate the fire safety analysis, the possiblaity or consequences of an accident not previously evaluated is not increased. The oiler does not impact RR pump operation. The oiler does not impact sny parameters or change any operating points thus, the margin of safety is not impacted.

TEMPORARY POWER TO AUXILIARY BUILDING RISER 1B AND FUEL BUILDING CRANE Document Evaluated: Temporary Modification 96 77 Log Number: %063 h Auxiliary Building Riser IB and the Fuel Building Crane are normally supplied by the 6.9KV bus IB through Unit Sub 10. Because of a 6.9KV bus IB outage, the Auxiliary Building Riser IB and the Fuel Building Crane were fed from the 6.9KV bus I A through Unit Sub IF. This change impacts Figure 8.3 1 of the USAR.

l h buses affected by this temporary modification are non safety and not required to furnish

, power under accident conditions. All equiprrent supplied by Unit Subs IF or 10 is non safety related. Thus, the probability of failure or consequences of failure of equipment important to safety are not increased. This temporary modification changes the source of non safety related power to non safety related loads, and does not impact any equipment or operating characteristics assumed in the accident analysis. Power connections were made on the non-safety portion of the

6.9KV electrical distribution system downstream of the Reserve Auxiliary Transformer (RAT).

, Therefore, the probability of occurrence or the consequences of any accident p' reviously evsluated l do not increase. The temporary modification provided breaker tap settings and breaker i coordination with upstream breakers. Wrefore, the possibility of occurrence or the

! consequences of accidents not previously evaluated did not increase. No safety related loads l were affected by this temporary modification. The load seen by the RAT was unchanged. As no j - safety related operating parameters or safety related setpoints were changed by this temporary

- modification, the margin of safety was not impacted.

SWITCH AND LIGHT REMOVAL FROM ECCS INJECTION CHECK VALVES Document Evaluated: Modification M-079 Partial Release # 6 Log Number: %-064 Modification M 079 removes the actuator hardware from the Emergency Core Cooling System

- (ECCS) testable check valves. The removal of the actuator hardware would prevent remote testing of IE12 F41 A, B, C, IE21.F006, and IE22-F005, h valve actuators, test push buttons and indicating lights allow partial stroke exercising the check valves during power operation. The modification was evaluated by Safety Evaluation Log Number 93 114, provided earlier in this L report. M-079 was to be lasted in five partial releases each corresponding to implementation of l - the modification on each valve. SE 93 ll4 evaluated M-079's intent ofissuance in five partial

releases, which could be implemented in any order, This partial release # 6 is an increase in the number of partial releases of M-079, Partial Release # 6 will delete the control room indication 4 and control room test push buttons for all five valves. This partial release will also charge the l

l

Attachment 2 to U 602836 Page $2 of142 main controf room mimic bus to reflect the removal of test push-button switches and indicating lights.

This partial release removes only the remote test and indicating features for the five ECCS check valves, and does not introduce any other hardware changes to the ECCS valves. The check valve actuator, and all support functions such as instrument air, limit switches, and manual actuators for aach individual check valve will remain in place to be modified by the respective partial release.

The analysis of SE 93 114 remains valid. This release does not remove the testable feature of the valves, as they can still be tested locally. The integrity of the ECCS check valves and the support functions is not affected by this partial change therefore, this partial release does not increase the probability or consequences of failure of the ECCS valves or any equipment important to safety.

Removal of the remote test feature and indication does not change the_ operation of these check valves and has no impact on the pressure isolation fbaction. Thus, partial release does not affect the probability or consequences of any accident previously evaluated in the accident analysis.

Similarly, removal of the remote test feature and indication does not introduce any new failure mechanisms or operating characteristics into the ECCS check valves and the possibility or . .

consequences of any new accident not previously evaluated is not increased. Removal of the remote test feature does not impact the p essure boundary fi-wtlon of the check valves and the margin of safety is not impacted.

. BLANK FLANGE INSTALLATION FOR RELIEF VALVE 1B21-F408 Document Evaluated: Temporary Modification 96-080 Log Number: 96-065

- This temporary modification installs a blank flange on the inlet Range to relief valve IB21-F408, Main Steam (MS) to Auxiliary Steam (AS). The relief valve provides ove.rpressure protection to the interface between the MS and AS system and is to be removed for refurbishment. In the current plant mode (cold shutdown), the MS system is not available as a pressure sourcs.

Temporary Modification 96-080 will be removed before'stanup.

USAR Section 9.5.9 statfs that the AS system is not safety-related. The AS system does not supply steam to any safety related component. Thus, failure of the blank flange would not increase the probability of failure or the consequence of failure of any equipment important to -

safety. The relief volve provides overpressure protection frc.m the MS system to the AS system.

USAR Section 15.6.4 postulates a main steam line break out:1le containment with the break being a guillotine break of one of the main steam lines. In tb current mode of operation, the MS

- system is not available and the accident analysis would not be applicable for a failure of the blank flange. Therefore, the probability of occurrence or consequences 'of the postulated steam line

. break outside containment would not be impacted by the temporary installation of the blind

~ flange. - As, the AS system is not considered in any technical specification, the margin of safety is not decreased.

Attachment 2 to U 602836

Page 53 of142 i j REFUSL PLATFORM UPGRADE TEST PROCEDURE (Fli 028) i i Document Evaluated

PAC No. 699-% for CPS Procedure 2800.62 R1 Log Number: 96-066  ;

This document is a change to a post modification test procedure written for Modification FH-028.

Modification FH-028 upgraded the control system of the refbeling bridge in containment. The upgrade installed a new Programmable Logic Controller (PLC) based control system and a I personal cornputer to interfhce with the PLC. Modification FH-028 also replaced the main hoist r right and left hand controllers, the position encoders, the main hoist load cells, and motor drive

units. Modification FH-028 was evaluated by Safety Evaluation Log Number 95 081.  ;

In order to install and align the new position encoders a temporary modification was installed to

' [

temporarily bypass three interlocks on the refueling platform in the containment. Temporary ,

Modification 95-041 installed thejumpers to allow movement of the crane so as to align the i

encoders. Temporary Modification 95-041 was evaluated by Safety Evaluation Log Number 95-087. Both Safety Evaluations,95-081 and 95-087, remain valid. This SE evaluates the change to CPS procedure 2800.62 R1 to incorporate the addition of two of the three jumpers added by Temporary Modification 95-04 L l

4 j This procedure change provides the administrative controls necest,ary for verification of the Installation of and removal of the temporaryjumpers. CPS procedure 2800.62 R1 does not allow 4

movement of any fuel bundles, moving the hoist belcw its normal elevation, and carrying loads

over the core. This procedure change does not negate or modify the prohibition against

, movement of fuel. Therefore, these changes to CPS procedure 2800,62 R1 will not increase the probability of failure or the conequences of fhilure of equipment important to safety. .

USAR Section 15.7.4 evaluates fuel handling accidents and USAR Section 15.4.1.1.2.2 discusses j rtid withdrawal error. Neither CPS proccuure 2800.62 R1 nor the PAC moves fuel assemblies. .

Similarly, neither CPS procedure 2800.62 R1 nor the PAC impact the rod block interlock. *

( Therefore, this procedure change has no impact on the accidents or their consequences previously j evaluated. This procedure change provides the procedural controls for the installation of the

. temporaryjumpers. SE 95-087 evaluated the addition of thejumpers and determined that they do >

4 not increase the possibility of a new and unanalyzed accident. This procedure change does not

! invalidate that conclusion.- Therefore, this procedure does not increase the probability or the consequences of a new type of accident not previously analyzed. The procedure change ensures i

that the temporary modification 95 041 is removed at the conclusion of the test. The procedure

- does not allow fuel movement and since the procedure change does not aff'ect any other interlock

. the procedure change does not impact the margin of safety.

DRYWELL FLOOR DRAIN OIL SEPARATOR AND WEIR BOX SCREENS Document Evaluated: Temporary Modification 97-003 Log Number: % -068 4

This temporary modification installs two screens in the Containment, Auxiliary, and Fuel Building 3_

Floor Drain Systems (RF) _ One of the screens is installed in the RF floor drain sump weir box, 1 - the other in the oil separator. The screens were installed to prevent foreign materhl from fouling

'. . = _ _ _ _ _ _ . , _ _ . _

Attaciunoid 2 to U 602836 Page 54 of 142 the weir box and the Drywell Floor Drain leakage monitoring instrumentation. This Temporary Modification chanoes the description of the drywell sump monitor in USAR Section 7.7.1.24.10.

The floor drain leak detection system is classified as not related to safety by USAR Table 7,1-13. -

Therefore, failura of the screens, which could affect the leakage detection instrumentation probe, would not increase the probability or the consequences of failure of equipment important to safety. There are no accidents evaluated in the USAR that pertain to the RF system and as such,

. the probability of accidents or the consequences of accidents is not increased This temporary .

modification does not affect leakage measurement, and does not impact the basis of any technical specification. Therefore, the margin of safety is unaffected.

RADIATION PROTECTION FACILITIES Document Evaluated: Changes to the USAR . Log Number: 96-069 This US AR change revises descriptions of several general use areas within the Radiological Controlled Area (RCA). This re-designation of general use areas reflects the merge of the Chemistry and Radiation Protection Departments. The changes include , conversion of the chemirtry storage area to an office space, clange of the non-contaminated laboratory to general technician area, removal of the mask drying ares, and a description of the current use of the laundry facility. These changes to the description of areas within the RCA do not affect technical specifications, plant, or operator responw to accidents or transient conditions, or system operation. ; These changes are consistent with Regulatory Guides 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," 8.8, "Information Relevant to Ensuring that Occupational Exposure at Nuclear Power Stations Will be As Low As Reasonably

' Achievable," and 8.10, " Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Reasonably Achievable," and do not affect any acceptance criteria of 10 CFR 19.12 or 10 CFR 20. These changes do not impact any systems or components related to safety and as such, do not affect the probability or consequences of any equipment failures or accidents, new or previously evaluated nor the margin of safety, ADD DRAIN LINES TO IB21-F032A & B CHECK VALVES Document Evaluaud: ECNs 29920 and 29921 Log Number: 96-071 ECN s 29920 and 29921 install %-inch ASME Class 2 drain valve assemblies to the check valve bodies in order to facilitate complete draining of the valves for Local Leak Rate Testing (LLRT)

. purposes. The drain assemblies consist of %-inch sc.edule 160 piping from the existing valve drain connection, a %-inch 1500 lb. ASME class 2 drain valve and a threaded pipe cap on the end of the assembly. These new drain valves are within the F032A & B valve containment boundary and become containment isolation valves themselves. The drain assemblies are attached to high-energy piping and are considered high-energy piping. USAR Section 3.6.2.1.4b exempts pipes of nominal size one-inch or less from pipe break evaluation and protection requirements. These valves would have no im> act on any postulated failure ofF032A & B. Thus, the addition of these -

drain assemblies would not increase the probability or consequence of failure of equipment related to safety. USAR Section 15.6 evaluates a feedwater line break, the break is assumed

Attachnunt 2 to U 602:36 Page 55 of142 instantaneous, circumferential, and downstream of the outermost isolation valve (10 feet downstream of the F032 valves). Failure of the drain assembly is bounded by this analysis. Thus, the addition of the drainlines has no impact on any of the accidents or the consequences of any accident previously evalented. Since the drain assemblies are constructed to meet ASME Boilec and Pressure Vessel code requirements, the possibility of occurrence or consequences of a new type of accident is not increased. The drain assemblies are designed and constructed to the same

- requirements as the F032 check valves, they will serve as containment boundary and will undergo the same test requirements as the check valves. Therefore, the margin of safety is not reduced.'

TEMPORARY MODIFICATION FOR BLIND FLANGES BETWEEN THE VENT PIPING OF THE RHR SUMPS AND THE AUXILIARY BUILDING HVAC SYSTEM (VA)

Document Evaluated: Temporary Modification 96-093 Log Number: % -072 This temporary modification installs blind flanges between the Residual Heat Removal (RH) pump room sumps vent pipes and the Auxiliary Building HVAC (VA) System in support of RF-6

activities requiring secondary containment. The connection between the sump vents lines and the VA system does not have isolation dampers. The connection of the RHR sumps to the VA exhaust ducts does not meet the secondary containment design guidelines of USAR Section 6.2.3.

The blind flanges will prevent venting the RH A, B, and C sumps to the VA system. The VA

. system is not safety related, and the blind flanges will have no impact on Rb sump operation.

Thus, the installation of the blind flarges will not increase the probability or consequences of failure of equipment important to safety. The addition of the Mind flanges will not change any safety system function or introduce any new operating parameter. The blind flanges will improve secondary containment integrity and the LOCA and Fuel handling accidents evaluated in the USAR will not be impacted. Thus, installation of the flanges will not increase the probability _ or consequences of any accident previously evaluated. The blind flanges do not introduce any new _

' failure mode. The possibility of a new type of accident or its consequences is not increased. The change eliminates a potential secondary containment leakage path, thus the margin of safety is not decreased.

REDUCTION OF REFUELING PLATFORM - DOWN TRAVEL CUTOFF MINIMUM SETTINGS Document Evaluated: ORM 4.6.3.d and CPS Procedures _ Log Number: %-073 '

9091.02 and 3703.01 This change revised the Refueling Platform main hoist mechanical down travel cutofflimit from "2-4 inches below the fuel assembly handle" to "l-4 inches below the fuel assembly handle." The revision was made because of historical difficulty in meeting minimum overtravel requirements.

- The cutofflimit restricts down travel of the grapple hook below the fuel assembly handle. This change revises a mechanical limit switch setting, which removes power from the main hoist to restrict its downward motion. The new limit switch setting does not introduce any new failure mode and thus, the probability or a consequence of failure of equipment important to safety is not impacted. USAR Section 15.7.4 evaluates a fuel handling accident that postulates dropping a fuel assembly _ onto fuel assemblies still in the vessel or stored in the spent fuel storage pool.

Attactunent 2 to U.602836 Page 56 of142 Procedural controls exist to preclude raining a fuel assembly without positive indication of the

. grapple fully engaged. Therefore, revising the minimum down travel mechanical cutoff will not increase the probability oflifting a fuel assembly without the grapple fully engaged sad the probability of occurrence or the consequences of a fuel' drop accident is not increar,ed. The mechanical interlock removes power from the main hoist to prevent further down travel, the interlock does not effect power to the mair. hoist when fuel is being moved. Therefore, the possibility of occurrence or consequences of an accident not previously evaluated are not increased. This revision to the mechanicallimit switch settings does not affbet fuel movement or the bases for any technical specification are not impacted.

INSTALL RELIEF VALVE 1CC318 ON CC HEADER ICC42D-6 IN CONTAINMENT (GL 96-06 ACTION).

Document Evaluated: ECN 29911 Log Number: 96-074 This ECN installs a relief valve ICC318 on the Component Cooling Water (CC) System header 1CC423-6 in containment. The relief valve is installed inboard of the inboard containment penetration valve ICC050, and is not within the containment penetration isolation boundary. The relief valve is set at 140 psig and will relieve to containment atmosphere if the header pressure exceeds the design pressure of 140 psig.

The relief valve was installed in response to Generic Letter (GL) %-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," concerns on potential overpressurization of containment penetration piping during containment heat-up following a postulated LOCA event. The relief valve assures that the pressure in the non-safety

. related piping remains within its design pressure and that non-safety related piping does not adversely affect the associated safety-related piping attacF ait.

The CC supply header in containment, line ICC42D-6, is non-safety related. The line is in the

- LOCA pool swell impact zone r.nd calculation IP-M-0362 evaluated the seismic and pool swell impact of the new relief valve. The addition of the relief valve will not result in structural failure of the relief valve or the header. Therefore, the addition of the reliefvalve will not result in an increase in probability or consequences of failure of equipment important to safety. The relief

. valve protects against overpressure due to thermal buildup during containment heat-up, as such

- the relief volume of the valve is small. The design basis LOCA bounds the effects postulated by a failure of this valvo. Therefore, the probability ofoccurrence or consequences of accidents previously evaluated are not increased. The CC header to which this pipe is installed is non-safety related and as such, does not introduce any new failure modes that could impact any safety function.~ Therefore, the possibility of or the consequence of any new type of accident not previously considered is not increased. The relief valve has been evaluated for various dynamic loads and found to have no impact on the CC piping. Containment integrity is preserved, as the valve is inside containment and inboard of tl.e containment penetration isolation boundary. The mr.rgin of safety is not reduced.

f

!. Attachment 2 to U-602:36 Page $7 of142 INSTALL RELIEF VALVE ICC319 ON CC HEADER ICC46B-8 IN CONTAINMENT (GL -

L

%-06 ACTION)

f. Document Evaluated: ECN 29912 Log Number: %-075 l This ECN installs a relief valve ICC319 on the Component Cooling Water (CC) System hes. der l ICC468-8 in containment; The relief vaive is installed inboard of the inboard containment j penetration valve ICCl28, and is not within the containment penetration isolation boundary. The relief valve is set at 140 psig and will relieve to containment atmosphere if the header pressure i

exceeds the design pressure of 140 psig. '

The relief valve was installed in response to Generic Letter (GL) %-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," concerns on potential overpressurization of containment penetration piping during containment heat-up

! - following a postulated LOCA event. The relief valve assures that the pressure in the non-safety i - related piping remains within its design pressure and that non-safety related piping does not

} adversely affect the associated safety-related piping attached to it.

The CC supply header in containment, line ICC46B-8, is non-safety related. The line is in the l LOCA pool swell impact zone and calculation IP-M-0362 evaluated the seismic and pool swell i impact of the new relief valve. The addition of the relief valve will not result in structural failure

= of the relief valve or the header. Therefore, the addition of the relief valve will not result in an l increase in probability or consequences of failure of equipment important to safety. The relief

valve protects against overpressure due to thermal buildup during containment heat-up, as such

, the relief volume of the valve is small. The design basis LOCA bounds the effects postulated by a l failure of this valve. Tierefore, the probability of occurrence or consequences of accidents i previously evaluated are not increased. The CC header to which this pipe is installed is non-safety j related and as such, does not introduce any new failure modes that could impact any safety function. Therdere, the possibility of or the consequence of any new type of accident not previously considered is not increased. The relief valve has been evaluated for various dynamic loads and found to have no impact on the CC piping. Containment integrity is preserved, as the valve is inside containment and inboard of the containment penetration isolation boundary. The margin of safetyis not reduced.

INSTALL RELIEF VALVE ICY 124 ON CY HEADER ICY 28C-6 IN CONTAINMEFTT (GL 96-06 ACTION)

Document Evaluated: ECN 29913 Log Number: 96-076 This ECN installed a relief valve ICY 124 on the Cycled Condensate (CY) System water header ICY 28C-6 in the Containment Building. The relief valve is installed inboard of the inboard containment penetration valve ICY 017, and is not within the containment penetration isolation boundary. The relief valve is set at 140 psig and will relieve to containment atmosphere if the header pressure exceeds the design pressure of 140 psig.

Attachment 2 to U 602836

, Page 58 of 142 The relief valve was installed in response to Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," concerns on potential overpressurization of containment penetration piping during containment heat-up following a postulated LOCA event. . The relief valve assures that the pressure in the non safety related piping remains within its design pressure and that non-safety related piping does not adversely affect the associated safety-related piping attached to it.

The CY supply header in containment, line ICY 28C-6, is non-safety related. The line is in the LOCA pool swell impact zone and calculation IP-M-0362 evaluated the seismic and pool swell impact of the new relief valve. The addition of the relief valve will not result in structural failure of the reliefvalve or the header. Therefore, the addition of the relief valve will not result in an increase in probability or consequences of failure of equipment important to safety. TF> relief valve protects against overpressure due to thermal buildup during containment heat-up, as such the relief volume of the valve is small. The design basis LOCA bounds the effects postulated by a failure of this valve. Therefore, the probability of occurrence or consequences of accidents previously evaluated are not increased. The CY header to which this pipe is installed is non-safety related and as such, does not introduce any new failure modes that could impact any safety function. Therefore, the possibility of or the consequence of any new type of accident not previously considered is not increased. The relief valve has been evaluated for various dynamic loads and found to have no impact on the CY piping. Containment integrity is preserved, as the valve is inside containment and inboard of the containment penetration isolation boundary. The margin of safety is not reduced.

INSTALL RUPTURE DISC IWX34M ON WX HEADER IWX12AA-2 IN CONTAINMENT

._(GL 96-06 ACTION)-

Document Evaluated: ECN 29915 Log Number: %-077 This ECN installs a rupture disc IWX34M on the Radwaste Reprocessing and Disposal (WX)

System header IWX12AA-2 in the Containment Building. The rupture disc is installed inboard of the inboard containment isolation valve IWX019, and is not within the containment penetration isolation boundary. The rupture disc has a burst pressure of 235 psig (nominal). This provides margin between the design pressure of the line and the rupture setpoint in order to prevent .

inadvertent rupture during normal operation. The rupture disc was installed in response to Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," concerns on potential overpressurization of containment penetration piping during containment heat-up following a postulated LOCA event.

The mpture disc assures that the pressure in the non-safety related piping remains within its design pressure and that non-safety related piping does not adversely affect the associated safety-related piping attached to it.

The WX effluent line from containment, line IWX12AA-2, is non-safety related. The line is in the containment steam tunnel and as such, is not susceptible to the effects of post-LOCA pool swell. Calculation IP-M-0362 evaluated the scismic loading on the rupture disc. Seismic / dynamic loading will not result in structural failure of the rupture disc, the header, or have any appreciable impact on the header. The addition of the rupture disc will not pose a potential missile hazard to a

Attachment 2 to U402836 Page 59 of142

. safety related equipment, Therefore, installation of the rupture disc will not result in an increase in probability or consequences of failure of equipment important to safety. The mpture disc protects the non-safety related WX piping against overpressure due to thermal buildup during containment heat-up following a LOCA, as such the relief volume, temperature, pressure, and potential radionuclides released are less than that assumed for the LOCA. The design basis LOCA bounds the efects of a mpture of the disc. The addition of the rupture disc does not adversely affect the probability or consequences of a LOCA. Therefore, the probability of occurrence or consequences of accidents previously evaluated are not increased. The WX effluent line, where the disc is installed, is non-saf(.ty related and has been evaluated for seismic

. occurrences. The rupture disc does not introduce any new failure modes, which could affect any safety function. Therefore, the possibility of or the consequences of any new type of accident not previously considered are not increased. The mpture disc has been evaluated for various dynamic loads during a seismic occurrence and found to have no impact on the WX effluent header. The burst pressure corresponds to the ANSI 150 lb. Class allowable working pressure for stainless steel piping. The mpture disc does not affect the basis for any technical specification and' therefore, the margin of safety is not reduced.

INSTALL RELIEF VALVE IWO327 ON WO HEADER IWOK3 A-6 IN CONTAINMENT

. (GL %-06 ACTION)

Document Evaluated: ECN 29917 Log Number: %-078 u This ECN installs a relief valve (lWO327) on the Plant Chilled Water (WO) System return header IWOK3 A-6 in the Containment Building. The relief valve is installed inboard of the inboard containment penetration valve IWOOO2B, and is not within the containment penetration isolation boundary. The relief valve is set at 250 psig and will relieve to containment atmosphere if the header pressure exceeds the design pressure of 250 psig.

The relief valve was installed in response to Generic Letter (GL) %-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," concerns on potential overpressurization of containment penetration piping during containment heat-up following a postulated LOCA event. The relief valve assures that the pressure in the non-safety related piping remains within its design pressure and that non-safety related piping does not adversely affect the associated safety-related piping attached to it.

The WO return header in containment, line IWOK3 A-6, is non-safety related. The line is not in the LOCA pool swell impact zone and calculation IP-M-0362 only evaluated the seignic loads on

' the valve for the header and nearby piping. The addition of the relief valve will not result in structural failure of the reliefvalve or the header. Therefore, the addition of the relief valve will not result in an increase in probability or consequences of failure of equipment important to safety. The relief valve protects against overpressure due to thermal buildup during containment heat-up, as such the relief volume of the valve is small. The design basis LOCA bounds the effects of a failure of this valve. Therefore, the probability of occurrence or a consequence of accidents previously evaluated is not increased. The WO header to which this valve is installed is non-safety related and as such, does not introduce any new failure modes which could impact any safety function. Therefore, the possibility of or the consequence of any new type of accident not

n Anachment 2 -

toU.6028% -

Page 60 of142 previously considered is not increased. The relief valve has been evaluated for various dynamic loads and found to have no impact on the WO piping. Containnent integrity is preserved, as the valve is inside containment and inboard of the containment penetration isolation boundary. The margin of safety is not reduced.

r CONTROL ROD DRIVE (CRD) REBUILD HEPA FILTER '

~ Document Evaluated: Temp Mod 96-092 ~ Log Number: 96-079 This temporary modification routes a suction hose and power cable to a 100 cfm HEPA filter in the Control Rod Drive (CRD) rebuild room through the ventilation system damper opening. The Fuel Building Ventilation (VF) System takes suction on the CRD rebuild room. The VF system

, exhaust 6 to the common station HVAC stack. On a loss of VF, the damper closes to prevent spread of contamination to clean areas of the fuel building. Routing ti e suction hose and power cable through the damper defeats the protection this damper provides spinst backflow and the spread of contamination. The spread of contamination on a loss of VF is not a concern since the HEPA filter was installed to filter air near the rebuild table thus limiting airborne contamination.

The damper is not used for Fuel Building isclation and does not provide a release path out of the secondary containment. The CRD rebuild room is a work area. Routing the suction hose and power cable through the damper will not increase.the probability or possibility of any accidents evaluated in the safety analysis. Neither the CRD rebuild room work area nor the ventilation damper affect any basis for any technical specification and thus, the margin of safety is not

- reduced.

JADD AIR CHAMBERS TO CONTAINMENT PENETRATION (IMC-52) PlPING TO LIMIT PRESSURIZATION-Document Evaluated: ECN 29909 LogNumber: 96-080 Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," identified that piping systems that penetrate containment may be susceptible to overpressure conditions due to thermal expansion of trapped fluids. This ECN adds an air volume tc existing Fuel Pool Cooling'and Cleanup (FC) System -

penetration,~ IMC-52, piping. This penetration has a capped line used for containment Local Leak Rate Testing (LLRT). This ECN changes the LLRT isolation valve from normally closed valve to normally open. This change adds piping off the line between the cap and the valve. - The piping has a two-inch air chamber and another LLRT isolation valve which will be in the normally closed position. This chamber will remain air-filled since there will be no flow (stagnant conditions) at this chamber location. As air is compressible, this will mitigate the effects of thermal expansion of water (i.e., overpressurization) # a postulated LOCA.

- This additional piping is constructed and tested to the same requirements as the existing piping Y and in accordance with USAR Table 3.2-1 requirements. As such, seismic, Quality Assurance (QA) requirements, safety classifications, etc. will be maintained. These chambers and associated piping perform a passive role of mitigating the potential for penetration overpressurization.

ee Attachment 2

,- to U.602836 Page 61 of142

Redundant motor operated containment isolation valves (CIVs) will still provide the active safety lt. function (isolation). Therefore, USAR Section 3.1.2.5.1, " Criterion 50 - Containment Design F Basis" equipment malfunctions or margin of safety would act be impacted by this change. As 4

passive items, these air chambers would not increase accident probability. The small size of the piping (1/2" & 3/4") connecting to the two-inch air chamber and the large (eight-inch diameter) j size of the FC penetration piping will prevent perturbations from the air chamber to affect the

j. '

penetration piping.- This change will subsequently not adversely affect the existing piping and not introduce a new type of accident or malfunction.

ADD AIR CHAMBERS TO CONTAINMENT PENETRATION (IMC-53) PIPING TO LIMIT PRESSURIZATION Document Evaluated:' ECN 29910 LogNumber: %-081 Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions l' identified that piping systems that penetrate containment may be susceptible to' overpressure conditions due to thermal expansion of trapped fluids. This ECN adds an air volume to existing Fuel Pool Cooling and Cleanup (FC) penetration, IMC-53, piping. This penetration has a capped line used for containment local leak rate testing (LLRT). This ECN changes the LLRT isolation valve from a normally closed valve to normally -

open. This change adds piping off the line between the cap and the valve. The piping has a two-inch air chamber and another LLRT isolation valve which will be in the normally closed position.

This chamber will remain air-filled since there will be no flow (stagnant conditions);at this chamber location. As air is compressible, this will mitigate the effects of thermal expansion of-water (i.e., overpressunzation) under a postulated LOCA.

This additional piping is constructed and tested to the e .e requirements as the existing piping and in accordance with US AR Table 3.2-1 requirements. As such, seismic, quality assurance (QA) requirements, safety classifientions, etc. will be maintained. These chambers and associated '

piping perform a passive role of mitigating the potential for penetration overpressurization.

Redundant motor operated containment isolation valves (CIVs) will still provide the active safety function (isolation). Therefore, USAR Section 3.1.2.5.1, " Criterion 50 - Containment Design Basis" equipment malfunctions or margin of safety would not be impacted by this change. As passive items, these air chambers would not increase accident probability, The small size of the piphs (1/2" & 3/4") connecting to the two-inch air chamber and the large (10-inch diameter) size

'of the FC penetration piping will prevent perturbations from the air chamber to affect the penetration piping. This change will subsequently not adversely affect the existing piping and not introduce a new type of accident or malfunction.

Attachment 2 to U 602836 Page 62 of 142 =

- CHANGE TO COMPONENT COOLING WATER (CC) SYSTEM OPERATION ,

I Document Evaluated:~ CPS Procedure 3203.01, TPD 96-0263 Log Number: %-082.

' This Temporary Procedure Deviation (TPD) provides guidance for manually controlling the Component Cooling Water (CC) System expansion tank level. The TPD is needed due to the automatic level control system being out of service (OOS). In addition, manual control of the CC expansion tank level le desired since automatic level control could mask a leak in the CC System.

This is contrary to USAR Section 9.2.2,2 which states that, "ari automatic drain valve prevents overfilling the tank. Makeup water to the tank is provided by an automatic level control system from the pretreatment and makeup domineralizer water system". This evaluation determines acceptability _of operating the CC expansion tank in this mode and makes USAR changes to allow - l this method ofoperation.

Failure of CC has been evaluated in USAR Section 9.2.2.3. This system is not required to assure safe shutdown of the plant. Therefore, this change will not impact design basis equipment malfunctions, failures or accidents. Since consequences of failure to the system are the same in either manual or automatic modes oflevel control, this change would not introduce new accidents or failures. The CC system's only mitigating function is containment isolation. This change would not impact this isolation function; therefore, this change will not reduce the margin of safety.

REMOVING BORON SAMPLING FROM RECYCLED WATER Document Evaluated: USAR Section 1L2,3 Log Number: 96-083 This change revises the USAR, Section 11.2.3.2, to eliminate boron from the recycled water sampling criteria.' Liquid waste processing at Clinton Power Station (CPS) results in two streams:

a product and a reject Stream. The product, or clean stream is normally returned to the primary cycle via the condensate storage tanks. The USAR requires that certain criteria be met to recycle water. The water quality requirements include li_mi's for conductivity, chlorides, boron, pH and radioactivity. There are three specific reasons that the boron analysis is unnecessary. There is a

- limited source of borated water at CPS. Secondly, processes are in place for removal of boron from process water such as radwaste demineralizers and condensate polishers. The third reason is in USAR Section 9.3.5.2 where it discusses that consequences of trace amounts (less than 50 ppm) of boron being inadvertently added to reactor water are minimal These first two reasons

< should limit boron concentrations to less than 50 ppm. USAR Section 9.2.6.1.1 states that the condensate storage system has no safety-related function except piping and valves forming containment isolation functions. This change would not impact the isolation functions.

Therefore, this change would not impact design basis equipment malfunctions. Design basis accidents evaluated in the USAR involve rapid increases in power from reactivity addition transients. Boron as a negative reactivity agent would decrease power should an inadvertent addition of boron occur. Removal of boron analysis of recycled water from a non-safety related system will not result in a new accident or malfunction since there is no uncontrolled source to contaminate this stream. Since other sampling criteria will maintain water quality, the Technical Specification Bases and margin of safety would be maintained.

- Anachment 2

. to U 602836 -.

Page 63 of142 GE RELOAD LICENSING ANALYSES FOR RELOAD-6, CYCLE 7 Document Evaluated: GE DRF J11-02920SRLR _ Log Number: 96-084

= and J11-02920 MAP GE report DRF J11-02920SRLR documents the results of reload licensing analyses for reload 6, cycle 7.1.ncluded in this document are the plant conditions assumed in the analyses, identified fuel bundle designs in the core and their proposed locations, applicable margin improvement and operating flexibility options, core and transient analysis results, and minimum power ratio (MCPR) operating limits. In J11-02920 MAP are maximum, average planar, linear heat -

generation rate (MPLHGR) values as a function of average planar exposure for each unique axial region (lattice) of the new fuel bundle to be irradiated for the first time in cycle-7; Before the start of cycle 7, the rea2or core will have been shuffled and 180 irradiated fuel bundles replaced by new GE10 bundles. Evaluation of GE10 fuel was reported in USAR Revision 5 (Log Number 93-0044). The new GE10 bundles loaded in cycle-7 will have the same mechanical and thermal-hydraulic design e .c _at two cycles while the nuclear design is slightly changed to meet cycle-7's energy requa s. The U-235 enrichment for this reload is 3.53%, which is slightly higher

- than the resider _ ales (3.22%,3.46% and 3.48%). Operating with these high discharge burnups required a change to the Safety Limit MCPR and an amendment to the CPS Operating License. This amendment (113) was approved by the NRC on January 22,1997.

' The mechanical and thermal-hydraulic design is the same as the last two cycies. Therefore, probability of a malfunction or creation of new accident types, related to this equipment, has not been changedi GE reanalyzed applicable USAR Chapter 15 accidents. The analysis showed that the original design basis consequences still bound the consequences due to this reload. Since the ' i reload meets the design and material standards as described in the USAR, there would not be an impact to design basis accident frequency or types. The core design change includes an operating

- limit change. The operating limit on MCPR was raised above the Safety Limit MCPR (SLMCPR) to accommodate the most limiting operating conditions and transients without exceeding the SLMCPR. Therefore, the margin of safety is maintained for this core configuration for normal, anticipated transients and event conditions.

CH' ANGE TO COMPONENT COOLING WATER (CC) SYSTEM OPERATION Document Evaluated: CPS Procedure 3203.01, TPD 96-0268 Log Number: 96-085 This temporary procedure deviation (TPD) is to provide improved instructions for use of the

. Shutdown Service Water (SX) System as a backup cooling water source to the Fuel Pool Cooling and Cleanup (FC) System heat exchangers (HXs). Normally, the FC HXs are cooled by the Component Cooling Water (CC) system. USAR Section 9.1.3, states that SX can be lined up to provide safety grade cooling water if CC is unavailable.

Valves ICC075A/B and 1CC076A/B are butterfly valves which are, at times, known to leak by their seats. This will affect the volume, pressure and cleanliness of the CC system as lake water enters the CC system through these valves. This would affect USAR Section 9.2.2.2 which states that the

. .~ -. . - - - . . . - - . - . ~ . - . - ~ . - - - - ~ . - - . - . - . - - . - - .-

t Attachment 2 E

to U-602836 Page 64 on42 domineralizer skid produces CC effluent quality of chloride less than 50 ppb and conductivity less than 3 mmho/cm at 25 'F. . This procedure is being revised to provide instructions to isolate the CC surge tank (to prevent over prenurization) and isolate the filter demineralizer in the event that CC is

. shutdown and SX must be used as the cooling water source while valves ICC075A/B and

, ICC076A/B are known ta leak.

j Failure of CC has been evaluated in USAR Section 9.2.2.3. This system is not required to assure

- a safe shutdown of the plant. The SK system is also being operated within the existing procedural limitations. Therefore, this change will not impact design basis equipment malfunctions,- failures or

- accidents. This change does not create a new type of accident or malfunctions as the known leakage will not impair the ability of the SX system to perform its safety function. The margin of safety

, would not be affected since CC is not needed for safe plant shutdown and the SX system is still operated within its design bases.

INSTALL RELIEF VALVE ON FP STANDPIPE (IFP47B-6) IN CONTAINMENT i

Document Evaluated: ECN 29958 LogNumber: 96-086

} GL 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," identified that piping systems that penetrate containment may be susceptible to

overpressure conditions due to thermal expansion of trapped fluids following a LOCA. This ECN j installs a reliefvalve on the containment Fire Protection (FP) standpipe. Installation of the relief j valve (IFPl%) will ensure that the pressure of the non-safety related piping remains within its design

pressure and tLat the non-safety piping will not adversely affect the associated safety-related piping attached to it. - This relief valve is installed within Containment Building and is not v/ithin the containment penetration isolation boundary. The relief valve will be set at 200 psig and will .

discharge to containment atmosphere if the FP standpipe header pressure exceeds the design pressure  :

l of 200 psi. This change affects USAR Figure 9.5 1, Sheet 9.

Failure of FP piping is described in USAR Section 9.5.1.2.2.3. The ECN evaluated design loads and the impact on structural missile hazard and FP header support loading. These evaluations determined that the change will not cause a line break of FP piping. These additions were qualified r to the same requiremerits as the existing piping. As such, there would not be an impact on design t

basis accidents or equipment malfunctions. The piping additions have been qualified to the same

- requirements as the existing piping and would therefore not create new types of accidents or

malfunctions. The FP standpipe is non-safety related and performs no direct function required to

. - safe shutdown. ~ The relief valve has been evaluatc! for various dynamic loads and will not fail in a -

! manner which would damage safety-related equipment in the area. Therefore, margin of safety

would not be impacted by this change.

h i

I

- . = . .- -- , - . _

Attachment 2 to U 602836 Page 65 of142

-INSTALL RELIEF VALVE ON FP STANDPIPE (1FP49B-6) IN CONTAIMMENT Document Evaluated: ECN 29959 LogNumber: 96-087 GL 96-06, " Assurance of Equipment Operability and Containment Intagrity During Design-Basis Accident Conditions," identified that piping systems that penetrate containment may be susceptible to overpressure conditions due to thermal expansion of trapped fluids following a LOCA. This ECN installs a relief valve on the containment Fire Protection (1 P) System standpipa. Installation of the relief valve (IFP197) will ensure that the pressure of the non-safety related piping twrains within its design pressure and that the non-safety piping will not adversely affect the associated safety-related piping attached to iti This relief valve is installed within the Containment Building and is not within the containment penetration isolation boundary. The relief valve will be set at 200 psig and will discharge to containment atmosphere if the FP standpipe header pressure exceeds the design pressure of 200 psi.' This change affects USAR Figure 9.5-1, Sheet 9.

Failure of FP piping is described in USAR Section 9.5.1.2.2.3. The ECN evaluated design loads and the impact on structural missile hazard and FP header support loading. These evaluations determined that the change will not cause a line break of FP piping. These additions were qualified to the sarr.e requirements as the existing piping. As such, there would not be an impact on design basis socidents or equipment malfunctions. The piping additions have been qualified to the same requiremeats as the existing piping and would therefore not create new types of accidents or malfunctions. ~ The FP standpipe is non-safety related and performs no direct function required to safe shutdown. The relief valve has been evaluated for various dynamic loads and will not fail in a manner which would damage safety-related equipment in the area. Therefore, margin of safety would not be impacted by this change.

INSTALL RELIEF VALVE ON FP STANDPIPE (IFP50B-6) IN CONTAINMENT Document Evaluated: ECN 29960 Log Number: 96-088 GL 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," identified that piping systems that penetrate containment may be susceptible to overpressurc conditions due to thermal expansion of trapped fluids following a LOCA. This ECN installs a relief valve on the containment Fire Protection (FP) System standpipe. Installation of the

. relief valve (IFP198) will ensure that the pressure othe non-safety related piping remains within its design pressure and that the non-safety piping will not adversely affect the associated safety-related piping attached to it. This relief valve is installed within the Containment Building and is not within the containment penetration isolation boundary The relief valve will be set at 200 psig and will dischange to containment atmosphere if the FP standpipe header pressure exceeds the design pressure

. of 200 psi This change affects USAR Figure 9.5-1, Sheet 9.

Failure of FP piping is described in US AR Section 9.5.1.2.2.3. The ECN evaluated design loads and the impact on structural missile hazard and FP header support loading. These evaluations determined that the change will not cause a line break of FP piping. These additions were qualified to the same requirements as the existing piping. As such, there would not be an impact on design basis accidents or equipment malfunctions. The piping additions have been qualified to the same requirements as the

Attachment 2 to U-602836

. Page 66 of 142

existing piping and would therefore not create new types of accidents or malfunctions. The FP standpipe is non-safety related and performs no direct function required to safe shutdown. The relief valve has been evaluated for various dynamic loads and will not fail in a manner which would damage safety-related equipment in the area. Therefore, margin of safety would not be impacted by this change.

ADD AIR CHAMBERS TO CONTAINMENT PENETRATION (IMC-56) PIPING TO LIMIT PRESSURIZATION Document Evaluated: ECN 29955 LogNumber: 96-089 GL 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," identified that piping systecas that penetrate containment may be susceptible to overpressure conditions due to thermal expansion of trapped fluids. This ECN adds an air volume to existing Fuel Pool Cooling and Cleanup (FC) System penetration, IMC-56, -

pipinb. This penetration has a capped line used for containment local leak rate testing (LLRT).

- This ECN changes the LLRT isolation valve from normally a closed valve to normally open. This change also adds piping off the line between the cap and the valve. The piping has a two-inch air chamber and another LLRT isolation valve which will be in the normally closed position. This chamber will remain air-filled since there will be no flow (stagnant conditions) at this chamber location. As air is compressible, this will mitigate the effects of thermal expansion of water (i.e.,

- overpressurization) under a postulated LOCA.

This additional piping is constructed and tested to the same requirements as the existing piping

. and is in accordance with USAR Table 3.2-1 requirements. As such, seismic, quality assurance

- (QA) requirements, safety classifications, etc. will be maintained. These chambers and associated piping perform a passive role of mitigating the potential for penetration overpressurization.

Redundant motor operated containment isolation valves (CIVs) will still provide the active safety function (isolation). Therefore, USAR Section 3.1.2.5.1, " Criterion 50 - Containment Design Basis" equipment malfunctions or margin on safety wculd not be impacted by this change. As passive items, these air chambers would not increase accident piobability, The small size of the piping (1/2" & 3/4") connecting to the two-inch air chamber and the large (10-inch diameter) size of the FP penetration piping will prevent perturbations from the air chamber to affect the penetration piping. This change will subsequently not adversely affect the existing piping and not

' introduce a new type of accident or malfunction.

ADD AIR CHAMBERS TO CONTAINMENT PENETRATION (IMC-82) PIPING TO LIMIT PRESSURIZATION Document Evaluated: ECN 29956 Log Number: 96-090 GL 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," identified that piping systems that penetrate containment may be susceptible to overpressure conditions due to thermal expansion of trapped fluids. This ECN adds an air volume to existing Fuel Pool Cooling and Cleanup (FC) System penetration, IMC-82, piping. This penetration has a capped line used for containment local leak rate testing (LLRT).

Attachment 2

- to U-602836 Page 67 of142

, This ECN changes the LLRT isolation valve firom a normally closed valve to normally open." This change also adds piping off the line between the cap and the valve. The piping has a two-inch dr chamber and another LLRT isolation valve which will be in the normally closed position. This chamber will remain air-filled since there will be no flow (stagnant conditions) at this chamber -

location. As air is compressible, this will mitigate the effects of thermal expansion of water (i.e.,

ovc2 pressurization) under a postulated LOCA.

This additional piping is constmeted and tested to the'same requirements as the existing piping and is in accordance with USAR Table 3.2-1 requirements. As such, seismic, quality assurance (QA) requirements, safety classifications, etc. will be maintained. These chambers and associated piping perform a passive role of mitigating the potential for penetration overpressurization.

Redundant motor operated containment isolation valves (CIVs) will still provide the active safety function (isolation). Therefc,re, USAR Section 3.1.2.5.1, " Criterion 50 - Containment Design Basis" equipment malfunctions or margin of safety would not be impacted by this change. As passive items, these air chambers would not increase accident probability, The small size of the piping (1/2" & 3/4") connecting to the two-inch air chan0er and the large (10-inch diameter) size of the FP penetration piping will prevent perturbations from the air chamber to affect the~

penetration piping. This change will subsequently not adversely affect the existing piping and not introduce a new type of accident or malfunction.

ADD AIR CHAMBERS TO CONTAINMENT PENETRATION (IMC-81) PIPING TO LIMIT PRESSURIZATION

--Document Evaluated: ECN 29957 Log Number: % -091 GL 96-06, " Assurance of Fquipment Operability and Containment Integrity During Design-Basis Accident Conditions," identified that piping systems that penetrate containment may be -

susceptible to overpressure conditions due to thermal expansion of trapped fluids. This ECN adds an air volume to existing Fuel Pool Cooling and Cleanup (FC) System penetration, IMC-81, piping. This penetration has a capped line used for containment localleak rate testing (LLRT).

This ECN changes the LLRT isolation valve from normally a closed valve to normally open. This change also adds piping off the line between the cap and the valve. The pip *mg has a two-inch air chamber and another LLRT isolation valve which will be in the normally closed position.- This chamber will remain air-filled since there will be no flow (stagnant conditions) at this chamber

. location. As air is compressible, this will mitigate the effects of thermal expansion of water (i.e.,

overpressurization) under a postulated LOCA.

This additional piping is constructed and tested to the same requirements as the existing piping and is in accordance with USAR Table 3.2-1 requirements. As such, seismic, quality assurance (QA) requirements, safety classifications, etc. will be maintained. These chambers and associated piping perform a passive role of mitigating the potential for penetration overpressurization.

Redundant motor operated containment isolation valves (CIVs) will still provide the active safety fimetion (isolation). Therefore, USAR Section 3.1.2.5.1, " Criterion 50 - Containment Design Basis" equipment malfunctions or margin of safety would not be impacted by this change. As passive items, these air chambers would not increase accident probabilit . The small size of the piping (1/2" & 3/4") connecting to the two-inch air chamber and the large (six-inch diameter) size a

l kttachment 2 to U-602836

Page 68 of 142 of the FP penetration piping will prevent perturbitions from the air chamber to affect the -

penetration piping. This change will subsequently not adversely affect the existing piping and not introduce a new type of accident or malfunction. -

REVISE ACCEPTANCE CRITERIA FOR EXCESS FLOW CHECK VALVES Document Evaluated: ECN 29982,29990-29997 Log Number: 96-092

~ Twenty-two (22) excess flow check valves (EFCVs) are affected by the changes made by ECNs 29982 and 29990 through 29997a Eight EFCVs have a lighter spring (lower spring constant)

- installed in the valve's poppet assembly so less differential pressure (and less flow) is needed to -

actuate the valve. The design specification of closing differential pressure and flow for all 22 EFCVs is being changed to coincide with the maximum pressure in containment following a design-basis LOCA. The remaining 14 EFCVs will not have hardware changes as they will fully meet the new specifications. These changes are corrective actions for condition report (CR) 1-96-10-406 The actions will bring the FFCVs into compliance with the design requirements within USAR Section 6.2.4.1 for compliance with Regulatory Guide 1.11, " Instrument Lines Penetrating Primary Reactor Containment."_ These valves were previously designed to close at a n

pressure higher than the peak accident pressure of containment. Consequently, these valves may not have been able to perform their safety function; actuating closed downstream of an instrument line loss during an accident, These ECNs require that the Technical Specification (TS) Bases and -

Operational Requirements Manual (ORM) be revised to reflect the revised acceptance criteria.

Two malfunctions of these EFCVs were evaluated in the USAR: Failure of the valves to close in the event of a rupture in the downstream instrument piping during an accident ud failure of the valve that would adversely affect the function of the connected instrument. These changes reduce impact of equipment malfunction since the lower spring constant will require less' differential ,

pressure (dp) to close and less containment leakage would be expected. rhis change is not t expected to have any detrimental effect on the EFCV failure mode or probability. Therefore,-

there would be no impact of the second malfunction. Failure of these high pressure EFCVs (ICM066 and ICM067)is evaluated in USAR Section 15.6.2. This evaluation concluded that these lines are small and terminate inside the engineered safety feature (ESF) fhration system and is bounded by the analysis for the most limiting fault (complete MSIV severance). At: such, this would not impact accidents evaluated as design basis. The configuration and function of the

- EFCVs remain the same. Therefore, there is not a potential to create a new accident type or

-_ equipment malfunction. As stated above, this change effectively decreases containment leakage during accident conditions. , As such, margin of safety would not be decreased.

MODIFICATION TO CONTAINMENT ISOLATION CHECK VALVES Document Evaluated: Modification FW-038 S1 Log Number: 96-093 Modification FW-038, Supplement I modifies the 1B21F032A/B outboard containment isolation check valves. The modifi::ations add a second actuator to improve closing characteristics. The spring assembly is eliminated from the existing actuator on each valve. This impacts USAR Section 5.4.9.3 and Table 6.2-47 Note 16. The valve shaft was also redesigned to improve the 4

Attachment 2 to U 602836 Page 69 of 142 closing characteristics. The second actuator is a mirror of the first having two air supply cylinders

_ per actuator. The spring assembly is no longer needed per me 16 cation FW-032, reported in -

USAR Revision 5 (Safety Evaluation Log Number 93-0091). In its place is a second air cylinder (four per actuator) to provide additional closing torque on the disc. The new shan design -

transmits the closing torque to the disc during the 'close' mode but will allow free movement of the shaR in the 'normalt mode. The current two-shaft design's replacement with a single shaR will reduce deflections and improve closing of the disc into the seat.

l These changes have been evaluated against the ASME Section III Class 1 valve stress analysis and existing actugt.or dynamic stress analysis and found to be acceptable. Therefore, these changes would not impact design basis malfunctions. Design basis accidents would also be unaffected by these changes since USAR Chapter 15 describes the feedwater (FW) line break and assumes one FW line break simultaneous with one main steam line break. _ The only plausible accident resulting from this modification is failure of FW piping outside this isolation valve. This is already postulated in the USAR. Margin of safety and primary containment would not be impacted since the modification qualifies the valve as before and ASME Code allowables are still met. The valves will be tested to ensure the LLRT acceptance rates are met.

- CPS EMERGENCY PLAN (E-PLAN), REVISION 10, ACN 11/3 Document Evaluated: E-Plan, ACN 11/3 LogNumber: 96-095 The Emergency Plan (E-Plan) was changed to update titles, building locations and renovations and to add other clarifications. These cunges did not decrease the effectiveness of the E-Plan per 10 CFR 50.54(q). Most chaages are minor in nature such as E-Plan Section 2.2.7 for the normal shift complement.- The changes also clarify that initial dose assessment is performed by Radiation Protection and takes credit for some of the enhancements provided by the rew version of MESOREM. This changed Symptom 3.1 for declaration of a Notification of Unusual Event (NOUE) and Alert by 43%. However, the new values are conservative by the guideline of NUREG-0654, ? Criteria for Preparation and Evaluation of Radiological Emergency Response -

Plans and Preparation in Support of Nuclear Power Plants." Symptom 4.1 was changed to make the procedure clear and easier to use. The section pertaining to d:!"s was also revised to reflect current NRC philosophy on the use of a utility exercise as a training session or drill. This ACN -

- does not affect the modes of failure or probability of malfunctions of equipment evaluated in the USAR. The early warning alarm still is in place. The NOUE is now at a higher level than this early warning alarm. This alarm will enhance emergency preparedness by giving operators early warning of an emergency classification that might be impending. These changes do not affect the nonnal operation of any plant systems or equipment and would therefore not involve an unreviewed safety question. This change does not involve the Technical Specification (TS) Bases nor the margin of safety.

Attachment 2 to U-602136

, Page 70 of142 CPS EMERGENCY CLASSIFICATION CHANGES Document Evaluated: EC-02, ACN 6/1 Log Number: 96-096 Within EC-02, " Emergency Classifications," the setpoints for Symptom 3.1 were changed from 7.0E-4 to 1.0E-3 and from 7.0E-3 to 1.0E-2 pCi/cc for Notification of Unusual Event (NOUE) and Alert respectively. The criteria for reactor shutdown category 7.0 were revised based upon changes to the Emergency Operating Procedures (EOPs). Symptom 4.1 for the NOUE wc '

revised to match the wording in the Technical Specification (TS) Limits. Also, a note was added for clarity stating that leakage is the sum ofidentified and unidentified leakage. This ACN does not affect the modes of failure or probability of malfunctions of equipment evaluated in the USAR. The early waming alarm still is in place. The NOUE is now at a higher level than this early warning alarm. This alarm will enhance emergency preparedness by giving operators early warning of an emergency classification that might be impending. These changes do not affect the normal operation of any plant systems or equipment and would therefare not involve an unreviewed cafety question. This change does not involve the TS Bases nor the margin of nfety.

TEMPORARY POWER TO FC DEMINERALIZATION OUTLET HEADER VALVE 1FC023 Demer.t svabsted: Temporary Modification 96-101 Log Number: 96-098 This Temporary Modification (TM) will provide temporary power to the Fuel Pool Cooling and Cleanup (FC) Demineralization outlet heater valve 1FC023 during the Division (Div) II bus outage. This is done in order to use the FC filter dcmineralizer (FD) during the Div II bus outage; 120V power from a non-divisional source will be used to supply control power to the valve. The power includes both Main Control Room (MCR) indication / control and solenoid power. This is contrary to USAR Section 8.3.1.4.2 as the valve will no longer be fed from a IE source.

The automatic LOCA function to close will not be active during the Div II outage. This is a result of the bus outage, not TM; however, the USAR states that this valve will isolate on a LOCA signal. It is acceptable per Technical Specification Table 3.3.6.1-1 for this LOCA signal to be inactive in MODES 1,2, and 3. When this TM is installed, the plant will be in a refueling mode (MODE 5). This TM will comply with Regulatory Guide 1.75, " Physical Independence of Electric Systems," in that it will not cause the tripping of any 1E circuits. Therefore, this TM would not affect equipment related to desip basis accident mitigation. This TM uses the same cable types and breakers; therefore, the probability of design basis accidents would not be increased. TS Table 3.3.5.1-1 states that primary containment and drywellisolation are only required in MODES 1, 2, and 3 as long as there is no operations with the potential to drain the vessel. During installation of this TM observing this requirement will ensure that the margin of safety is maintained.

l

oAttachment 2 ,

to U402836 -

3 ,

Pape 71 of142 IMPROVED TECHNICAL SPECIFICATION (ITS) BASES CHANGE FOR EOC-RPT .

- Document Evaluated: BE 96-011 LogNumber: 96-099

This Improved Technical Specification (ITS) Bases charige revises the End-of-Cycle Recirculation 4

Pump Trip (EOC-RPT) assumed breaker interruption time and mechanical timing acceptance ,

criteria. The overall assumed time is revised from 95 milliseconds to 80 milliseconds. The basis

for the 80 milliseconds is the 60 milliseconds design specification interruption time of the RPT l breakers plus 20 milliseconds load driver, signal propagatior, and additional margin time. The l' - RPT mechanical timing acceptance criteria is revised from s 34 milliseconds to s 41 milliseconds -

; for trip coil 2 (TC2) and s 34 milliseconds for TCl. CPS uses 4 Westinghouse type DVP circuit breakers to power two reactor recirculation (RR) pump motors. These breakers have two trip -

coils each, TC1 and TC2. Either will trip the breaker with TC) having inputs from the breaker

[ control circuitry and protective relays.' TC2 only has one trip input from the Nuclear System

! Protection Systeni (NSPS) System which is associated with the ITS 3.3.4.1 function of EOC RPT

+

instrumentation. This ITS Bases change is in accordance with ITS 5.5.11, " Technical

Specification Bases Control Program."

i- These changes to the ITS Bases do not impact the timing requirements of any ITS safety limits I

associated with EOC-RPT functions < Mechanical timing testing at the specified acceptance i criteria decreases the probability of equipment malfunctions because it provides additional

! assurance that the breakers are operating properly. This change is within that recommended by

! the vendor. Therefore, this change would not create new accidents. This change wili maintain l the operation of this equipment as it is designed and would therefore not impact accidents evaluated as design basis. Changing the timing values will maintain the safety analysis of the ITS L Bases and thus would not impact margin of safety, j CONTAINMENT /DRYWELL ISOLATION VALVE OPERABILITY i.

! Document Evaluated: CPS Procedure 9064.02 LogNumber: %-100 l GL 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis L Accident Conditions," identified that piping systems that penetrate the containment may be L susceptible to overpressure conditions due to the thermal expansion of fluid that is heated. It is postulated that containment isolation could trap water between the inboard and outboard isolation valves. ..As the containment heats up, following a LOCA, the trapped water would heat up end expand, potentially overpressurizing the piping segment. The GL %-06 resiew identified that containment penetrations IMC-078 and IMC-088 and Standby Liquid Control (SC) System penetration IMC-116 were identified as potentially susceptible penetrations. A new procedure, i CPS procedure 9064.02, "GL 96-06 Containment Penetration Drained Verification," will be g ' developed to identify proper valve lineup and method to drain piping between isolation valves.

The Component Cooling Water (CC) System is not required to assure safe shutdown of the plant

' and the change would therefore not affect design basis equipment malfunctions or accidents. The safety function of the SC penetration is to maintain isolation during accident conditions. This line l

d Attachnent 2 41 to U.6028% - ,

, Page 72 of142 a

v is capped in the Fuel Building. As such, there would be no effect on the CC or SC system .

L operation; therefore, there is no effect on the probability or consequences of an accident or equipment malfimetion.- Draining there penetrations will support the design basis function of the penetrations; containment integrity, Changing these valves from open to closed will not initiate a new accident or equipment malfunction since the design function is to close for containment i

3  : integrity This change maintains containment integrity and therefore, does not decrease the

! margin of safety. ,

l DRYWELL FLOOR DRAIN INLEAKAGE MONITORING SYSTEM Document Evaluated: Modification LD-028 Log Number: 96-101 R1 l( Modification LD-028 is to measure unidentified dre .31 floor drain leakage rate with close to real l' tirae response characteristics. This modification wili Se a redundant system to the two existing j monitoring syAems. The other systems arc weir box and LD-027 and Process / Display Computer

[ (CX/CZ) calculations, it will be added to the Technical Specification (TS) Bases Section B3,4,7, The ler.kage detection system provides the necessary sensors and instrumentation used to indicate

< abnormal leakage from the reactor pressure vessel (RPV), and in certain cases, initiate signals

- used for automatic closure ofisolation valves extenal to the Drywell. LD-028 will be part of the pnmary leak detection method used to detect small unidentified leaks. - This modification will measure the_Drywell floor drain sump level using a bubbler type level detection system. The resulting level signal will be fed to a FLC located ir he main control room (MCR) and the signal

!. converted from level change to inleakage amoun .he leakage rate will be continuously indicated j and/or recorded in the MCR. Before the unident ied leakage increases to a tomi flow rate of five

.. gpm or increases by at least two spm within a 24-hour period (TS limits), the detecting -

! instrumentation will trip and actuate an alarm in the MCR, This change will not result in an

isolation trip. This change will enhance the secondary detection method (LD-027) used to detect 3

gross unidentified leakage. The enhanecments include more efficiently writing the code to calculate the two gpm increase within 24-hour alarm, adding a wide range for larger flow rates (0- _

64 gpm) and adding a regression calculation for decreasing le.kage rates at low flows.

Regulatory Guide 1,45, " Reactor Coolant Pressure Boundary Leakage Detection Systems,"

implies that only one system is necessary to fulfill the design bas's for this system.. As such, failure of this system would still not affect the other two systems in place and design basis equipment malfunctions would not be impacted.' Accidents evaluated in the USAR do not take credit for leakage detection.- Therefore, eis change would not impact design basis accidents. Failure of any component for this modification would not affect RPV integrity, The piping and tubing are designed to the original architect /eng:neer criteria which would preclude the potential for this equipment creating new types of accidents or equipment malfunctions. There is no impact on the margin of safety since there is no credit given for this system in any accident type.

L - , .

s Attaciuneet 2

- te U 602836

_ Page 73 of142 CLARIFICATION ON USE OF SHELL TYPE ANCHORS AT CPS Document Evaluated: USAR Change for CR 1.%12-086 Log Number: 97-001

- and ECN 29989 This USAR change to Section 3.9.1.1 clarifies the exclusion on shell type anchor bolts.

In NRC Bulletin 79-02, " Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts," Rev. 2, the NRC identitled design and installation deficiencies on pipe support base plates that use concrete expansion anchor bolts for Seismic Category I systems. In the bulletin, the

- NRC required licensees to initiate a testing program to assure minimum design requirements were

met relative to cyclic loading and embedment depth, unless them' formation could be verified via Quality Control documentation. - In response to the bulletin, CPS initiated a test program but excluded shell type anchor bolts on the basis that they were not used at CPS. This USAR change t

modifies CPS's response to the bulletin by clarifying that shell type anchors are not used in Seismic Category 1 applications at CPS. This change to the_USAR and the NRC Bulletin 79-02 '

response was prompted by station condition reports that identified use of shell type anchor bolts in non-seismic category 1 applicationsc Shell type anchor bolts will be limited to non-seismic category I applications and used on a case by case basis.

- USAR Section 3.2.1 discusscs design of non-seismic category I structures, cystems, and components that could impact seismic category I structures, systems, and co.nponents. By their exclusion from seismic category 1 applications, and by compliance with the requirements of

- USAR Section 3.2.1, the use of shell type anchor bolts at CPS will not increase the probability of occurrence or the consequences of failure of equipment important to safety. USAR Section 3.2.1 states that non-seismic category 1 structures, whose collapse could impact safety related components, ate designed and analytically checked to confirm their integrity against collapse when subjected to scismic loading resulting from a safe shutdown earthquake. Therefore, the use of shell type anchor bohs will not increase the probability of occurrence or the consequences of any accident previously analyzed.

Failure of the anchor shell type anchor bolt would not cause the non-seismic category I structure to impact safety related equipment. Therefore, the use of shell type anchor bolts would rot increase the possibility of or the consequences of an accident not previously analyzed. The use of shell type anchor bohs in non-seismic category 1 applications does not impact any Technical Specification and the margin of safety is not impceted. 4 CONDUCT OF OPERATIONS / UPDATED SAFETY ANALYSIS REPORT Document Evaluated: USAR Chapter 13, CPS Forms 1401.01 F001 Log Number: 97-002 through F009 4

This change to USAR Chapter 13 modifies position titles, position descriptions, and organizational descriptions. Additionally this change deleted discussion of the Radwaste Operator and the Radwaste Operations Center (ROC) operator as these positions were eliminated. The Rad 3vaste Operator position was merged with the area operator position, and the ROC operator position was re-designated as the Unit attendant. This change to Chapter 13 of the USAR is y ,

l Attachment 2

- to U-@2836

, Page 74 of 142 j solely administrative and does not impact any structures, systems, or components important to safety. The administrative change does not impact the accident analysis or the margin of safety.

DRILL HOLES 1N DISC TO LIMIT PRESSURIZATION DUE TO LOCA HEATUP Document Evaluated: ECNs 29906,29907,29908,29919, and 29929 Log Number: 97-003

. In Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," the NRC identified that piping systems that penetrate containment may be susceptible to thermal expr.nsion and overpressure during heatup following a LOCA. In the GL the NRC postulated that water trapped between the inboard and outboard

- containment isolation valves would expand post LOCA. The above listed ECNs resulted in a 1/8-inch hole being drilled in the inboard disc of the following valves: 1CC050,1CC053, ICY 017, 1CC057, and ICCl28. The 1/8-inch hole provides a leakage path to relieve pressure due to thermal expansion of trapped fluid following a LOCA. As fluid between the containment isolation valves thermally expands, the outboard face of the inboard disc will be pushed offits seat and the pressure will be relieved through the hole in the opposite disc face. CPS calculations show that the 1/8-inch hole will preclude the penetration assemblies from experiencing significant pressure buildup or increases. Leak Rate Testing will verify that leakage past the 1/8 inch hole will meet the required leakage guidelines.

The 1/8-inch hole drilled into the disc does not pose any seismic concerns, and as the expected leakage past the valve is expected to be small, the valve will continue to function as a gate valve.

Therefore, the probability of occurrence or consequences of a failure of equipment important to safety will not be impacted. The inner disc will remain seated when the LLRT pressure is applied and the containment isolation function is still provided by the redundant motor operated containment isolation valves. Therefore, the probability of occurrence or the consequences of any accident previously evaluated do not increase. The 1/8-inch hole does not introduce any new failure modes, thus, the possibility of occurrence or the consequences of an accident not previously considered are not increased The 1/8 inch hole prevents failure of the penetration assembly and maintains the design basis. The containment leak rate tests to verify the leakage is below the maximum allowed will meet the requirements of Technical Specification 3.6.1.1; therefore, the margin of safety is not impacted.

TEMPORARY POWER TO DG MU HTG MCC 1 (1 AP87E) DURING UNIT SUB IE BUS OUTAGE Document Evaluated: Temporary Modification (MWR D73885) Log Number: 97-004 The normal power supply to Diesel Generator Make Up (MU) Heating Motor Control Center (MCC) 1 (1 AP87E) is Unit Sub IE (I AP15E) cubicle 4B. Because of a Unit Sub IE outage, a temporary power feed from Unit Sub IM was provided to keep 1 AP87E in service. This temporary power feed affects the configuration shown in USAR Figure 8.3-11. The connected

- load of 1 AP87E,253 amperes, will not adversely affect the operation of Unit Sub IM or of the auxiliary power system.

I a

Attachment 2 to U-602836 l Page 75 of 142 Unit Sub IE and Unit Sub IM are both non-safety related, the loads normally supplied by Unit  :

Sub IE are also non safety related.- The loads normally supplied by Unit Sub IE will temporarily l i

be supplied by another non-safety-related source. Therefore, the probability of failure or the consequences of failure of equipment important to safety are not increased. The temporary power feed to I AP87E does not impact the accident analysis, therefore the probability of occurrence or the consequences of an accident previously analyzed.will not incream. Both Unit Sub IE and Unit Sub IM are supplied by the non-safety 6.9KV Buses. The connected load of1 AP87E does not have an adverse effect on Unit Sub IM because the load added by 1 AP87E is within the capacity of the Unit Sub. The temporary power feed does not introduce any new failure modes, therefore the change will not impact the possibility of occurrence or the consequences of an accident not previously evaluated. As all the loads affected by this temporary power feed are non-safety and are not considered in the basis for any technical specification; therefore, the margin of safety is not impacted.

PLANT CHANGE FOR DRYWELL EQUIPMENT DRAIN SUMP FLOAT SETPOINTS Document Evaluated: ECN 29769 Log Number: 97-005 This ECN changes the setpoints for the mechanical alternator and the high-high level switch settings for the Drywell Equipment Drain sump. This ECN also changed the calibration data reference point. This reference point is used in setting the mechanical alternator setpoint. The reference point was changed from the bottom of the sump pit to the top of the cucb. Finally, this ECN removes reference to the 25 gpm alarm in the Main Control Room from the USAR.

The mechanical alternator and high-high level switch setpoints were lowered by 3 inches. 'IrJ:

change was made to minimize indication errors caused by wuter backing up into the weir box before the sump pump started. The change in reference point for taking data facilitates field calibration of the instrument and only changes the reference point for taking the data. Lastly, since the pump start setpoint was changed, the decreased volume in the sump will affect the sump fill timer cycle and the pump mn timer cycle. The number of pump runs will increase while the total pump run time will remain the same. Since the 25 gpm annunciation is calculated from the sump fill cycle and the pump run cycle, the annunciation will occur at a point different from 25 gpm. The annunciation will be a general alarm that indicates abnormal flow rates.

Neither the change in setpoint, nor the reference point used for adjusting the setpoint, nor removal of the 25 gpm reference to flow will impact the probability or consequences of failure of I equipment important to safety. The drywell equipment drain sump flow monitoring instrumentation is non-safety related and does not affect the accident analysis. Therefore, the probability of occurrence or consequences of accidents previously evaluated are not increased.

These changes do not introduce any new failure modes or operating points important to safety, thus, the possibility of, or the consequences of an accident not previously evaluated is not increased. The change in operating level for the drywell equipment sump does not affect l Technical Specification 3.4.7 or the basis for that technical specification. Therefore, the margin of safety is not reduced.

Attaciunent 2 to U402836 Page 76 of 142 BLIND FLANGE INSTALLED FOR RUPTURE DISC 1WX34M Document Evaluated: Temporary Modification 97-007 Log Number: 97-006 L

ECN 29915 resulted in a rupture disc (lWX34M) being installed on the Radwaste Reprocessing and Disposal (WX) System header IWX12AA-2 in the containment steam tunnel. The rupture disc was installed in response to GL E06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," concerns on potential overpressurization ofcontainment penetration piping during containment heat-up following a postulated LOCA event. The installation of IWX34M was evaluated by Safety Evaluation (SE) 1 Log Number %077, reported earlier in thh report. This temporary modification replaced IWX34M with a blank flange. The temporary blank flange was installed because of a rupture disc rupture that occurred on January 9,1997. This temporary modification will be removed before exiting MODES 4 and 5 when a permanent rupture disc design is complete and installed. The temporary blank flange affects the design configuration presented in USAR Figure 11.4-3. The blank flange is a temporary and passive mechanical device to preserve the system pressure boundary.

. SE %077 evaluated the addition of the rupture disc to this non-safety-related piping. The '

seismic and dynamic analyses performed for the rupture disc remain valid for the flange.

Therefore, as with the rupture disc, no safety related equipment will be affected by the addition of 9 L the flange and the probability or consequences of failure of equipment important to safety are not increased The temporary blank flange will be installed only during MODES 4 and 5 and the overpressure protection required by the GL will be provided by another rupture disc before entering MODE 3. The integrity of the WX piping is maintained, and no credible new and not previously evaluated accidents have been created. Therefore, the effects of overpressute due to thermal expansion following a LOCA will not increase. The temporary addition of this blank flange to the WX piping does not affect the basis for any technical specifications and the margin of safety is not decreased.

INSTALL RUPTURE DISC AND RELIEF VALVE ON WX HEADER IWX12AA-2 (GL

%06)

Document Evaluated: ECN 30052 Log Number: 97-007 ECN 29915 resulted in a rupture disc (lWX34M) being installed on the Radwaste Reprocessing and Disposal (WX) System header IWX12AA-2 in the contairunent steam tunnel. The rupture disc was installed in response to GL %06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions,", concerns on potential overpressurization of containment penetration piping during containment hea:-up following a postulated LOCA event. The installation of IWX34M was evaluated by Safety Evaluation (SE)

Log Number %-077, reported earlier in this report. Temporary Modification 97-007 replaced lWX34M with a blank flange. The Temporary Modification was evaluated by SE Log Number 97-006 also reported earlier in this report. The temporary blank flange was installed because of a rupture disc rupture on January 9,1997. The disc failure occurred during flushing operations following a radwaste transfer. The disc installed by ECN 29915 had a set pressure of 235 psig.

Attachment 2 to U-602836 Page 77 of 142 The disc failure was attributed to a number of cumulative effects, which decreased the design margin.- ECN 30052 specifies a rupture disc with a set pressure of 500 psig. A relief _ valve was also installed to mitigate installation errors of the disc. Additionally, a pressure gauge was installed between the disc and the relief valve to monitor the condition of the disc. The rupture disc of ECN 30052 replaces the disc of ECN 29915 and as such, is installed in the same location and constructed of the same insterial as the disc of ECN 29915. Calculation IP-M-0362 analyzed the seismic and dynamic loading of the disc / relief valve assembly and the analysis and calculation are similar to that performed for the failed disc. As with the disc of ECN 29915, pool swell impact is not a concern due to the location of the installation. The seismic and dynamic loading is the same for both discs. The new disc / relief valve / pressure gage assembly weighs approximately 20 pounds, and its effect on the header and on the piping analysis is similar to the rupture disc installed under ECN 29915. The analysis of SE 96-077 remains vslid and the conclusion that there is no impact on the probability and consequences of failure of equipment related to safety remairs the same. The analysis of SE Log Number %-077 of a design basis LOCA bounding the rupture of the disc, installed under ECN 29915, remains valid for the new rupture disc / relief valve /preswre gauge assembly. Therefore, the probability or consequences of a design basis accident it. not increased. The new assembly, like the previous rupture disc, does not introduce any new failure modes and as such, the pouibility or consequences of a new type of accident is not increased. The installation of the new assembly, to replace the previous rupture disc, does not impact any Technical Specification Bases and thus the margin of safety is not decreased CHANGE TO DIESEL GENERATOR OVERSPEED TRIP SETPOINT RANGE Document Evaluated: ECN 29951 Log Number: 97-008 ECN 29951 changed the overspeed trip range for the Division I, II and III diesel generators. The new range of 1015 to 1045 rpm will standardize the overspeed trip ranges on all diesel generators,

= will minimize unnecessary inadvertent trips of the diesel generators and will provide the necessary protection against overspeed conditions. USAR .5 tion 8.3.1.2.2 specifies the speed that the generator may not exceed on a single largest load rejection.- The changes in overspeed trip range affect this section.

The diesel generato 3 have a nominal speed of 900 rpm, and are designed for sustained operation

-- at 1050 rpm. The change in trip range of this ECN is not outside the design parameters of the diesel generater and as such does not increase the probability offailure of equipment important to safety. Similarly, as the new overspeed trip range is within the design parameters of the diesel generator, and less than the deign for sustained operation speed of the generator, the new range will minimize inadvertent diesel generator trips and will not increase the probability and

~

consequences of an accident previously analyzed. As the new range is within the design prameters and does not introduce any r.ew failure mode, the new range will not increase the possibility or consequences of any new type of accident not previously analyzed _ The new range will not affect any basis for technical specifications and as such, the margin of safety will not be affected.

Attachment 2 to U-602836 l- Page 78 of 142 -

' 3D MONICORE DATABANK CHANGE FOR CYCLE Document Evaluated: ECN 29558 Log Number: 97-009-This ECN implements the Cycle-7 new databank files on the 3D Monicore Core Monitoring Computer (C91-P646). The 3D Monicore is a non-safety system used to ensure compliance with the limits in the operating license manual and monitor coce performance. The data bank contains

- Cycle-7 licensing information, design data, and as-loaded core data that is based on previously-approved licensing information. The data in the databank ls CPS Cycle-7 specific, consistent with or bounded by the approved licensing documents and will not ir. crease the probability of or the comequences of failure of any equipment important to safety.

The data in the databank is consistent with the Cycle-7 core design and as such will ensure that 1

- the initial conditions and operating characteristics assumed as initial conditions in the accident analysis are maintained. The information in the databank is not used in any active control -

function. Therefore, the information in the databank will not increase the probabilitp of'an accident or the consequences of an accident previously evaluated. The information in the databank does not introduce any new failure modes or operating characteristics consistent with the ore design and as such, will not increase the possibility or conseqdences of accidents not previously evaluated. Because the information in the databank is bounded by the approved licensing documents and reflects the core design data, the margin of safety is not reduced.

RELAXATION OF REQUIREMENTS DURING INTERLOCK TESTING Document Evaluated: ORM 2.6.7, Mode Switch Position Log Number: 97-010 This change revises Operational Requirements Manual (ORM) 2.6.7 " Mode Switch Position" to lock the reactor mode switch in the refuel or shutdown position except as provided in the special t - operations section of the Technical Specifications. The change allows the Re .ctor Mode Switch to be unlocked while performing Reactor Mode Switch Interlock Testing using station procedures. This change allows the administrative controls of the ORM to match the special operations technical requirements. This change, to allow interlock testing, will not increase the probability of or consequences of failure of equipment important to safety as the requirement to lock the mode switch is only relaxed while the limiting conditions for operation of Technical Specifications 3.10.2 and 3,10.8 are in effect. The controls of LCO 3.10.2, in effect during interlock testing, require th .t all rods are fully inserted and no core alterations are in progress.

USAR Section 15.4 discuues rod withdrawal error at low power. The administrative controls of the LCO preclude a rod withdrawal error; therefore, the probability of occurrence or the consequences of a Rod Withdrawal Error are not increased Operation of the mode switch as allowed by LCO 3.10.2 and ORM 2.6.7 will not introduce any new failure mechanisms thus, the possibility of a new accident is not created. The Technical Specification bases allow for testing of the interlocks if the requirements of LCO 3.10.2 are met.

j The margin of safety is not reduced by allowing the mode switch interlock to remain unlocked.

l

Anachment 2 to U402836 Page 79 of 142 ELIMINATE CONTROL ROOM VENTILATION (VC) SMOKE MODE DURING CORE ALTERATIONS Document Evaluated: ORM 2.4.3 " Control Room Ventilation Log Number: 97-011

- Smoke Mode" This change to the Operational Requirements Manual (ORM) 2.4.3 " Control Room Ventilation S . soke Mode" operational requirements eliminates Core Alterations as an applicable mode for-operability. Amendment No. I12 to the CPS Operating License incorporated a new special operations LCO 3.10.10 " Single Control Rod Withdrawal-Refueling." The amendment allowed

- control rod withdrawal in MODE 5 without requiring the secondary containment and main control room boundary to be established. This change to the ORM eliminates a requirement that conflicts with the amendment. The Main Control Room HVAC (VC) smoke node is a standby mode of operation for the VC system. This change does not alter the mode of operation er affect any hardware in the plant. As such, this change does not increase the probability of or consequences of failure of equipment important to safety. Operation of the VC system in the smoke mode is not an initiating event for an accident evaluated in the USAR. As such, this change does not affect the probability or consequences of accidents previously evaluated.

Operation of the VC system in the smoke mode does not introduce any new failure modes; therefore, this change does not increase the possibility of any new type of accident not previously evaluated. This change is consistent with the new Technical Specification 3.10.10 and will not redace the margin of safety.

REVISION TO USAR FIGURE 9.5-1 DUE TO ECN 30054 Document Evaluated ECN 30054, USAR Figure 9.5-1 Log Number: 97-012 ECN 30054 resulted in a flanged connection being installed on Fire Protection (FP) line 0FP02D.

The flanged connection was installed to accommodate.an offset due to piping shift. The flanged connection was installed in accordance with approved station documents. The flanged connection was installed in the FP ring header. The flow characteristics or requirements of the ring header

=

are not affected. The operation of the FP pump and its flow characteristics are unaffected. The addition of this flanged connection does not affect any safety-related equipment and as such, the probability or consequences of failure of equipment important to safety are not increased.

The connection does not affect the ability of the FP system to deliver the required flow and

- perform its design function. Therefore, the probability or consequences of an accident previously evaluated are not increased Addition of the flanged connection will not impair the r.bility to achieve or maintain a safe shutdown thus, the margin of safety is not reduced.

ADD MANUAL FLOW CONTROL VALVE DOWNSTREAM OF GENERATOR STATOR COOLING SYSTEM (GC) COOLERS Document Evaluated: ECN 29825 Log Number: 97-013

i, - ,

j. Anachment 2 ys ,

to U-602:36 Page 80 of142 i~ Thir ECN resulted in a manual flow control valve, a flow element, and an indicator being added to -,

the Service Water (WS) System downstream of the Generator Stator Cooling (GC) System L . coolers. Additionally, a test vent used for data collection is being removed. The addition of the U valve will preclude having to throttle flow with the existing buttedly valves. The flow element n and indicator are added to allow setting the required flow. This change affects USAR Figurc 9.2-l :1, The changes are being done on the non-safety, non-seismic WS system. The addition of the  ;

j valve, flow element, and indicator does not affect the GC system as the chan8es are installed on ,

the WS side of the cooler. The installation of the valve, flow element, and indicator, was done to

[*

. the same requirements as the origind piping. The change will not affect the probability or.

consequences of failure of equipment important to safety. The WS system is not required for safe

!- shutdown of the plant. The valve will not act as initiator for any of the accidents evaluated in the

accident analysis. Therefore, the probability of an accident or the consequences of an accident

[ previously evaluated will not increase. The valve, flow element, and indicator will not introduce i any new failure modes and as such, will not increase the possibility of a new type of accident not b previously evaluated. The GC or WS systems do not affect any technical specification bases and I as such the margin of safety is not uduced.

CATHODIC PROTECTION OF DIESEL GENERATOR FUEL OIL LINES Document Evaluated: USAR 1.8, RG 1.137 LogNumber: 97-019 In August 1995, a fire protection line leak was attributed in part to insufficient cathodic protection. As part of the root cause determination, an extensive cathodic protection survey was conducted. . The survey determined that some areas near buildings had low pipe-to-soil potentials as compared to National Association of Corrosion engineers (NACE) standard RP-01-69,

" Recommended Practice - Control of External Corrosion on Underground or Submerged Metallic Piping Systems." USAR Section 1.8 states that CPS is committed to Regulatory Guide (RG) 1.137 " Fuel Oil Systems for Standby Diesel Generators." RG 1.137 states that buried portions of the fuel oil system should be provided with protective coatings and cathodic protection in accordance with NACE RP-01-69. Not all underground fuel oil system piping meet the criteria of NACE RP-01-69. Safety Evaluation (SE) 97 019 cvaluates a USAR change that would allow the -

Diesel Fuel Oil fill lines not to meet these criteria. All other com,oonents of the Diesel Generator Fuel Oil and Transfer System comply with NACE RF-01-69. Lack of cathodic protection on the oil fill lines would not impair the ability to fill the storage tanks. Water in-leakage into the storage tanks would not prevent the diesel generators from performing their funcion. Water in the storage tanks would rettle to the bottom and be removed by a drain at the bottom of tank when performing Technical Specification (TS) surveillance SR 3.8.3. Therefore, lack of cathodic protection would not increase the probability or consequences of failure of equipment important to safety.

USAR Table 9.5-2 Item 5 analyzes a loss of the diesel generator fuel oil fill line. Tl" loss of the fill line would not p*eclude supplying fuel to the diesel generator during prolonged periods of operation since refueling can be accomplished using a spare nozzle on the storage tanks.

Therefore, the loss of the fill line would not cause loss of the diesel generator and would not preclude safe shutdown following a design basis LOCA and a loss of offsite power. Therefore, the lack of cathodic protection would not increase the probability or consequences of any accident

Attachment 2 to U 602836.

Page 81 of142 previously evaluated in the safety analysis._ Lack of cathodic protection on the fill line does not introduce any type of new failure mode than previously evaluated and thus, will not create any new type of credible accident. The possibility or consequences of any new type of accident'are not increased. TS B'ases section B3.8.3 requires that each diesel generator be provided with a seven-day oil supply. Lack of cathodic protection will not deprive the diesel generators of this

- fuel supply. Therefore, the lack of cathodic protection will not reduce the margin of safety.

-INSTALL ANTI-SWEAT FLEXIBLE ELASTOMERIC INSULATION ON LINES 1RF26E AND 1WX12B (GL96-06)

Document Evaluated: ECN 29904 and 29905 Log Number: 97-020

. ECNs 29904 and 29905 resulted in anti sweat flexible elastomeric insulation being installed on lines IRF26E and IWX12Bi This change is in response to GL 96-06 " Assurance of Equipment Operability and Containment Integriy D, iring Design-Basis Acci.ier,t Conditions." GL 96-06 postulated overpressurization of containment penetration pipirg due to thermal expansion during a heatup following a design basis LOCA. The insulation wi's reduce the heat transfer rate into the penetration piping during a post design basis LOCA heatup thus maintaining containment penetration integrity, This change er.sures the plant's ability to meet the requirements of USAR Section 3.1.2.5.1. The insulation is buoyant and if fails and falls int'o the suppression pool it will not clog the Emergency Core Cooling System (ECCS) suction strainers. Thus, the addition of the insulation will not increase the probability or consequences of failure of equipment important to safety, The insulation is not an initiator for any design basis accidents previously evaluated. The insulation will reduce the heat transfer rate into the penetration piping during a heatup following a -

LOCA thus preserving containment penetration integrity; therefore, the addition of the insulation will not increase the probability or consequences of any accident previously evaluated in the USAR. The insulation will not introduce any new failure modes or operating characteristics, therefore, the possibility or cotuequences of any new type of accident not previously evaluated will not increa:e. The insulation will not compromise any basis for any technical specifications;

, therefore, the margin of safety is not decreced.

INSTALL RELIEF VALVES 1RF092,1RE049, AND IMC087 ON NON-SAFETY RELATED PIPINGIN CONTAINMENT (GL96-06) ,

Document Evaluated: ECNs 29914,29916,29918, and 29976 LogNumber: 97-021

- These ECNs resulted in relief valves 1RF092,1RE049, and IMC087 being installed on non-safety related piping in the Containment Building. The relief valves are installed inboard of the inboard containment penetration valves IRF021, IRE 021, and IMC008, and are not within the comainment penetration isolation boundary, The relief valves are set at.the respective system design pressures and will relieve to the containment atmosphere if the header pressure exceeds the

' design pressure.-

The relief valves were installed in response to GL %-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," concerns on potential

overpressurization of containment penetration piping during containment heat-up following a

Anhm 2 to U402sw Page 82 of142 postulated LOCA event. The relief valves assure that the pressure in the non safety related piping remains within its design pressure and that non safety related piping does not adversely affect the associated safety related piping attached to it.

Calculation IP M-0362 evaluated the seismic and pool swellimpact of the new relief valves, h addition of the relief valves will not result in structural failure of the relief valves or the headers to which they are attached. Therefore, the addition of the relief valves will not result in an increase in probability or consequences of failure of equipment important to Wety. The relief valves protect against overpressure due to thermal expansion of fluid during containment heat up

- following a design basis LOCA, as such the relief volume of the valves is small. The LOCA evaluated in the USAR as a design basis accident bounds the postulated volume, temperature, and pressure discharged by these valves. Therefore, the probability of occurrence or consequences of_ _

accidents previously evaluated are not increased. The headers to which these relief valves are installed are non safety related and as such, the installation of the relief valves does not introduce any new failure modes that could impact any safety function. Therefore, the poasibility of or the consequence of any new type of accident not previously considered is not increased The relief valves have been evaluated for various dynamic loads and found to have no impact on the ,

respective piping to which they are attached. Containment integrity is piesert xi, as the valves are  !

inalde containment and inboard of the containment penetrathn iso 4 tion boundary. Therefoie, margin of safety is not reduced. l SAFETY EVALUATION FOR USAR CHAPTER 15, DROPPED FUEL CASK ANALYSIS Document Evaluated: USAR 15.7.5.3.3 LogNumber: 97-022 Safety Evaluation (SE)97-022 evaluates a USAR change made in 1994. CR l 96-12-046 identified that the subject USAR change should have had a full safety evaluation performed. The change to USAR Section 15.7.5.3.3 clarified the discussica of the efects of a dropped fbel cask on the 737 foot level of the Fuel Building railroad bay 'I h s USAR change resulted in text being changed to indicate that although the slab would not fail, the umlerside could spall, causing pieces of concrete to fall on safety-related equipment below. The USAR previously stated that there was no safety related equipment which could be affected by a cask drop accident. The USAR was changed to state, "The underside of the concrete slab could spall causing pieces ofconcrete to fall on safety-related equipment below. Procedural restrictions are in place to ensure safe handling of the cask."

This USAR change made no physical changes to the plant. The effect of a dropped cask on r:earby safety related equipment was previously evaluated and documents referteced in CPS Supplemental SER 5 and plant procedures remain in effect to ensure that the operation of hoisting a spent fuel cask remains safe. These procedures ensure compliance with NUREG-0612,

" Control of Heavy Loads at Nuclear Power Plants, Resolution of Generic Technical Activity A-6." The rystems that could be affected by a dropped cask accident were reviewed. The review

. determined that redundant equipment would be available if any spalled concrete effected underlying equipment. As such, there is no change in design basis equipment malfunction or

- accident. The drop and spalling werJ evaluated and it was determined that this would not create a

- = = = - ==.

---._..----.-.------- -- _.~.- - ..--

Attadunent 2 4

to U 602436

, Page 23 of142 new design basis accident or equipment malfunction. Review of the Technical Specification Bases

] for any systems involved, determined that the change would not impact margin of safety.

CRD REBUILD ROOM EQUIPMENT REPLACEMENT l

Document Evaluated: ECNs 29113,29859,29879 LogNumber: 97-023 i

' These ECNs repuhod in replacement mechanical equipment being installed in the Control Rod l

Drive (CRD) rebuild room. This equipment replaces equipment that was removed as a result of  ;

ECN 28942 (USAR Revision 6, SE Log Number 95-010). The new equipment consist: of a L CRD proflush tank with pump and Alters, ultrasonic tank with pump and filter and various ancillary equipment for rebuilding and/or cleaning the CRDs. The new equipment provides a more efficient and lower dose method of rebuilding CRDs. This change impacts USAR Figure i >

12.3 50 due to the addition ofl# art radiation shielding around some of the installed equipment. This change was evr.luated for seismic and structuralimpact. The change would not l Impact rebuilding CRDs as this is done in accordance with CPS procedures that specify rebuild  ;

i requirements. None of the equipment replaced in this room interacts with any safety related i equipment. Therefore, this change would not impact design basis accidents, malfunctions or

, create new design baals accidents. None of this equipment could impact safety related equipment, j therefore margin of safety would not be reduced. ,

CYCLE-SPECIFIC HYDROGEN GENERATION INFORMATION INCLUDED IN USAR Document Enluated: USAR Sections 1.8,6.2.5 Log Number: 97-024 l

This change adds cycle-specific information regarding hydrogen generation following a LOCA into the USAR. The change also includes USAR cross-references to cycle-specific values, updates the analysis results curves with the latest results, describes the conservatism in the analysis so that recalculation will not be necessary each cycle and corrects the hydrogen I concentration acceptance criteria to wet conditions as evaluated in Safety Evaluation (SE) i Number 93-0038 and reported in USAR Revision 5. USAR Section 1.8 to Regulatory Guide 1.7,

" Control of Combustible Gas Concentrations in Containment Following a Loss-of Coolant Accident," was also revised to remove the portion that states the USAR contains input and output

-values from the original FSAR. The analyzed hydrogen concentrations remain below the combustibility threshold of 4%. Therefore, potential increased equipment malfunction due to hydrogen combustion is not present. Although the change slightly increases the quantity of metal

. - that can react to post accident hydrogen, the limiting concentration is not being changed. As

[ such, probability of an accident, increases in frequency of accidents or malfunctions have not been changed or a new type of accident has not been created. The margin of safety as defined in the

- bases for any Technical Specification (TS)is not affected. TS Bases B3.6.2.1 addresses the

- function of the hydrogen recombiners to protect the containment from a potent. ' breach cwsed by a hydrogen-oxygen reaction. The recombiners which are intended to maintain the hydrogen concentration below 4% are not affected by this change.

W-TpwVi F77 U-- '-

F9Yy 't-rg W---'z??eM w w w m 7- *ry p rg F W-Weenw m-m'- 11W?e1v 9y'-M*i ---**w- '

Anachment 2

- to V402sw Page s4 or142 ALLOWANCE FOR USE OF KOOLPHEN K TYPE INSULATION Documer.t Evaluated: ECN 30017 LogNumber: 97 027 This ECN allows the use of Koolphen K phenolic foam type insulation for low temperature Off Gas (VO) HVAC ducts and equipment. The original design specilled fiberglass insulation.

USAR Appendix E, Section 4 is being revised to add this as an insulation type allowed to be used on low temperature piping. HVAC duct insulation is non safety related and performr, no safety function. This insulation has better thermal conductivity properties than the existing insulation (fiberglass) and does not absoit water due to its rigid cell structure. The insulatlan weighs less than fiberglass. W material meets the flame spread and smoke criteria limits as specified in USAR Appendix E. Calculations have determined that this material will not increase the fire load classifications in the affected zones. This material is being used on the VO HVAC duct and equipment. W VO HVAC system does not perform any safety function; therefore, design basis equipment malfunctions are not impacted. This insulation is not being used in containment.

Therefore, USAR Section 6.2.2.2 regarding material entering the suppression pool is not affected along with USAR accidents. As stated earlier, this material has better lasulating properties and is acceptable from fire protection and loading concerns. Therefore, this material would not create new accidents or equipment malfunctions. Insulation, or its need, is not assumed in the bases for the Tecialcal Specifications; therefore, margin of safety would not be affected by using this insulation.

s USAR CHANGE TO INCORPORATE CURRENT CYCLE RELOAD ANALYSIS RESULTS Document Evaluated: USAR Change Package 7 203 Log Number: 97-028 CR l-96-08-115 was written when it was determined that CPS had not included cycle-specific information explicitly in the USAR. Instead, a reference to the cycle specific Supplemental Reload Licensing Report (SRLR) has been made in the USAR following each refueling outage, The specific concem a$ stated in the CR is that the referenced information is not easily accessible to anyone requiring this infonnation in support of a safety evaluation. The corrective action plan for the CR requires a revision of the USAR to incorporate the applicable cycle-specific information for cycle 7 and remove leformation not applicable to the current cycle including reference to previous SRLRs. This revision consists of the creation of a new appendix to Chapter 15 of the USAR which will contain the cycle-specific information from the cunent SRLR as well as references to this new appendix la appropriate sections of the USAR. ,

These changes and the a#ition of US AR Appendix 15D are to make the necessary changes to the USAR to provide the proper docuntentation in support for the current cycle core design. The current cycle core design wus evaluated under Safety Evaluation (SE) %-084, reported earlier is this revision to the USAR. As such, these changes have already been reviewed and an unreviewed safety question was determined not to exist,

__ _ = _ _ =--

Awh 2

. to U4028%

Page t$ of142 OPERABILITY OF PENETRATIONS DUE TO CIV LEAKAGE (GL 96-%)

Document Evaluated: CR l 9610 360-OD LogNumber: 97 029 GL %-06, " Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions," identified that piping systems that penetrate the containment may be susceptible to overpressure conditions due to the thermal expansion of fluid that is heated. It is

_ postulated that containment isolation could trap water between the inboard and outboard Containment Isolation Valves (CIVs). As the containment heats up, following a LOCA, the trapped water vrould heat up and expand, potentially overpressurizing the piping segment. This evaluation is to determine that operability of containment is maintained to support NRC ASME Code overpressurization requirements thereby satisf'ying USAR design basis requirements of 3.1.2.5.1 and 6.2.4.1. To support penetration operability, a two step approach is being implemented for certain penetrations which have known leakage rates as determined by the local leak rate test (LLRT) Program. This safety evaluation (SE) is the first step for determining that current leakage is sufficient for ovsrpressurization protection. This step involves no physical change to the containment barrier. The second step, not addressed in this SE, is to add overpressure protection during the next refueling outage (RF-7). The aff'ected penetrations are IMC-050,065,069,070,103 and 104 which include the Makeup Condensate Storage System (MC), Solid Radwaste Reprocessing and Disposal System (WX), Fuel Building Equipment Drains System (RE), Containment, Auxiliary, and Fuel Building Floor Drain System (RF), and Chilled Water System (WO) Systems.

Analysis (IP-M-0425 R/0) for these penetrations shows that currently measured CIV leakages are sufficient to prevent overpressurization and loss of the design basis function of containment integrity. This analysis demonstrated that the containment stmeture and its internal compartments can accommodate, without exceeding the design leakage rate, with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA. Therefore, design basis

. malfunctions and barrier / margin of safety are not aff'ected by temporarily using leakage to justify operability. This is an additional method of performing the function of overpressurization protection using the existing configuration, Since these penetrations will continue to perform their intended function, this will not increase the probability or consequence of accidents previously evaluated in the USAR. Individual leakage rates could change; therefore, permanent changes are scheduled, via Maintenance Work Requests (MWRs) for RF-7, to install more direct methode of relieving potential overpressures. To ensure that maintenance activities do not change the leakage amounts, MWR tags were hung in close proximity to these valves to alert maintenance personnel. These actions are sufficient until RF-7 in order to ensure that there is no potential for creating new types of accidents or equipment malfunctions, BYPASS LINE AROUND CC HEAT EXCHANGER (ILX)

, Document Evaluated: ECN 23904, Modification WS-024 LogNumber: 97-030 This ECN/ modification resulted in a bypass line with a manually operated control valve being installed on the Plant Service Water (WS) line around the Component Coolireg Water (CC)

System heat exchangers (HXs). This will allow flow back to the lake, thereby lowering WS

_ __ _ =

_..m.

l l

i Auacionent 2 l J

lo U402836 Page 86 of142

header pressure which will reduce cavitation problems throughout the system. These component are shown in USAR Figure 9.2 1. The WS system is non safety, non-seismic. The bypass line is i

designed and built to the same standards as the existing WS piping is. The bypass line supports

, are designed for postulated accident conditions. Therefore, this line would not be postulated to 1

fail and impact safety related equipment. For the same reason, since there is no impact to safety- i

! related equipment. Since the change was made to the non safety, non seismic WS system, there is -

d no impact on design basis accidents. Since the bypass is constructed to the same standard as the other portions of the WS system, the change would not introduce potential unanalyzed accidents.

- The change has no impact on any system, structure, or component assumed in the basis for any technical specification; therefore, margin of safety would not be impacted by this change. ,

5 TEMPORARY BLANK FLANGES ON FP RING HEADER Document Evaluated: TM 97-013 Log Number: 97 031 l i Piping stresses have caused Fire Protection (FP) line OFP02D to shiR approximately 1/2 inch.

l This shift caused the cast iron valve body of 0FP106 to crack. Temporary Modification (TM) 97-l 013 installed blank flanges la place of FP ring header valve 0FP106. Line OFP93 A is the Service

- Building FP line of the ring header. This line is in the vicinity of 0FP106. This TM also removed

valve 0FP108 from line 0FP93A and installed a blank flange on the ring header side. This TM

allowed repairs to the damage caused by the stresses and shlRing while allowing the jockey fire pump vice the diesel-driven fire pump to keep the ring header pressurized. This change temporarily interrupted fite protection water to the Service Building. However, compensatory

measures were taken per CPS procedure 1893.01, " Fire Protection Impairment Reporting."

Having the FP system in service in this configuration is contrary to USAR Figure 9.5 1, Sheet 2.

The flanges are located at a position where the FP system could be isolated if a leak were to i develop. The blank flange results in the same effect as closing the valve and icolating this section

! - of piping. Therefore, this change dou not affect the ability of providing water to the FP system

. and the reactor, and therefore, would not increase probability, consequences of accident or

_ equipment malfunctions defined as design basis. Allowing the FP system to be returned to service l would not create new accidents. This change has no impact on the function of the FP system j- except for the Service Building supply. Isolation of the Service Building is allowed and was

- compensated for in accordance with CPS procedure 1893.01.- As such, the margin of safety l assumed in the basis for any technical specification would not be reduced.  ;

REVISION TO USAR ATTACHMENT D3.6 FIGURES i- Document Evaluated: CR l %-12169 Log Number: 97-032 i

h _

Condition Report (CR) 1-96-12-169 identified discrepancies in pipe whip restraint information

shown on USAR figures in Attachment D3.6, " Summary of Failure Mode Analysis for Pipe .

Breaks and Cracks," in that the figures were out-of-date. CR investigation determined that data

' in Attachment B3.6, " Dynamic Effects of Postulated Pipe Ruptures," is correct and reflects the as-built condition. This evaluation is to add a note to 14 discrepant figures in Attachment D3.6 to

- refer users to the correct pipe whip restraint data in B3.6. The note will state, "Do not use this W4me m .w.- , , - , . . . -r-., w e m - c r---# - -. , , + - ..w,.,-m,, , b 4,w -- - - ---m----,--n%rqe --w w , f - , - - , w.,-vv-- ,- - ,, ,,---wn+---

Atemetianame 2 m U402sw Page 87 of142 Agure for pipe whip restraint Information. Please refer to USAR Attachment B3.6 for acairste information on this subject." This note supplements the text of USAR Attachment D3.6 that already directs the roadw to B3.6 for pipe whip restraint data. USAR Attachment B3.6 fbily evaluated the possibility and consequence of cracking and failure of high energy lines and identilles necessary whip restraints. As such, this clarincation would not impact design basis malfbnction or accidents. This change does not physically change any plant equipment or methods of operation. Therefore, the change would not create the potential for new accidents or malAmotions of equipment important to safety. Since the capability of the pipe whip restraints has not been reduced, the margin of safety is not affected by this change.

REVISION TO USAR SECTION 13.3, EMERGENCY PLANNING Document Evaluated: USAR Section 13.3 LogNumber: 97 034 This is an administrative change which revises USAR Sect!on 13.3, " Emergency Planning," to remove unnecessary details which are contained in the Emergency Plan (E Plan). These changes did not change the E Plan. Therefore, the change did not decrease the effectiveness of the E-Plan per 10 CFR 50.54(q). Since this change does not change the E Plan or the way which it is implemented, all requirements for emergency classifications, operation under emergency -

conditions and actions taken to mitigate accidents would be unchanged. As such, changes to design basis accidents, malfunctions and margin of safety would be unchanged.

REACTOR OPERATOR (RO) TRAINING PROGRAM DESCRIPTION, REVISION 8 Document Evaluated: RO Training Description, Rev 8. LogNumber: 97-035 Training Course CN87022, "RO Generic Fundamental Review," was in ti e original draA of revision 7 to the Reactor Operator (RO) Training Program Description. Prior to its issuance, it was decided not to add this course. However, all references were not removed prior to revision 7 issuance, When the RO Training Program Description, Revision 8 was draRed, references to CN87022 were removed. This also added a matrix of training requirements for CN86311," Fuel Handling Training." This does not require a USAR, Operational Requirements Maisual or Technical Specifications change. The RO Training Program Description is an administrative procedure. Changes to this administrative procedure, i.e., correcting of references, would not change design basis accidents, malfunctions and margin of safety.

REVISE INACCURATE DESIGN STATEMENT, CR l-96-07-060 f

Document Evaluated: USAR Section 15.6.2 -

LogNumber: 97-036 USAR Section 15.6.2 contains an inaccurate statement that no instrument or sample lines are

- connected to the reactor coolant pressure boundary which penetrate the primary containment.

This statement is immediately followed by a correct statement about the instrument and sample lines. The CPS design drawings verified that the second statement was correct, that there are two instrument lines and two post accident sample lines connected to the reactor coolant system

.which penetrate primary containment. The USAR change will replace the incorrect wording with

= , =-

. ==

Anaciuncat 2 -

to U4028%

Page as o(142  ;

a reference to LaAR Section 15.6.4, Steam Line Break, which is bounding on Instrument Line i Breaks. All four lines and associated valves for the penetrations meet applicable design criteria.

As such, design basis malfbnctions of this equipment would not be affected. Instmment line breaks are bounded by main steam line breaks. Therefore, design basis accidents would not be i affected by this correction to an inaccurate USAR statement. No new accidents or equipment malfunctions are created since the instrument lines have been designed and analyzed in accordance ,

with Regulatory Guide 1.11 " Instrument Lines Penetrating Primary Reactor Containment." Since  ;

the failures associated with this change are bounded by more limiting failures, the acceptance limits are unaffected, therefore margin of s.afety is not reduced.

CHANGE TO USABLE CAPACITY OF CONDENSATE POLISHERS Document Evaluated: CPS Procedure 6801.01 Log Number: 97-038 The USAR currently states that condensate polisher resin beds will be replaced when the calculated resin capacity approaches 50% ofinitial capacity. This change to CPS procedure 6801.01, " Capacity of Condensate Polishers," will permit the use of condensate polisher resin with a capacity less than 50% provided, the reactor is shutdown, feedwater is not in service to the reactor and the Circulating Water (CW) System is not in service. This is a change to USAR Section 1.8, Regulatory Guide 1.56, " Maintenance of Water Purity in Boiling Water Reactors."

This change will reduce the amount of radioactive waste produced and reduce unnecessary resin  ;

bed exchanges.

The basis for the 50% capacity margin is to ensure adequate capacity is available to permit an orderly shu.down in the event of serious inleakage of condenser cooling water before reactor water quality standards are exceeded by potentially harmful constituents oflake water. During shutdown conditions, Feedwater (FW) and CW are not in service and the additional resin margin would not be not needed. Therefore, there is no increase in design basis equipment malfunction introduced by this change. Prior to placing the reactor in service, the polisher capacity will r.eet the 50% capacity criteria. _ As such, probability or consequences of a design basis accident would not be irueased. Since there is not a path for lake water to enter the reactor using the condensate system during shutdown conditions, this change would not create a new accident type not already

- evaluated in the USAR. Since reactor water cannot be contaminated by lake water, no new equipment malfunction will occur and this change will not reduce the margin of safety as defined

- in the basis for any technical specification.

REPLACE RELIEF VALVES WITH THOSE OF DIFFERENT VENDOR Document Evaluatedf ECN 29848 Log Number: 97-039 ECN 29848 allows the use of an alternate valve as the low and high pressure foodwater heater channel and drain cooler channel relief valves. This change requires a revision to USAR Figure 10.4-9 to reflect use of Dresser relief valves as equivalent replacements for Crosby relief valves.

These valves are classified as non-safety-related, ASME Section VIII valves per USAR Table 3.2-

- 1. The operating characteristics of the replacement Dresser valves are the same as the original valves except for the capacity on valves IDV006A/B, IDV014A/B, IDV024A/B, IDV044A/B, 4

ousbe m rus-u e'mi-rrw nae-me?Mw-  % p -'wg-e--e _a3yame-w--------&e-i-s'-

yy y be ywwy v m griry-*ww r -ve

-r- qur+e-1 y v-+' y g---y p-- - =w-g- vF'N1l{W" W "- e+7 T r+ -'+ F ('

Anaciument 2 to U 602836 Pape 89 of142 IDV053 A/B and IDV059A/B. W Dresser valves have a slightly larger orince and thereby higher capacity. The increased orifice / capacity has a negligible effect on the fbnction of the system since the discharge line is routed to open equipment drains having the capability of handling the additional flow.

There are no equipment malfbnctions evaluated in the USAR for this relief system. Therefore, this change to non safety equipment would not change design basis equipmt.nt malfbnctions. The loss of feedwater heating in USAR Section 15.1.1.1 is not affected by the change since this event is not affected by a relief valve change. As such, the change does not impact design basis accidents. Since the relief valves are determined to be equivalent in function, having identical set pressures, this change would not introduce new failures, malfunctions or accidents. Since USAR Section 15.1.1.1 is not impacted, barrier integrity and margin of safety would not be affected.

PR F018 TEST PROCEDURE Document Evaluated: CPS Frocedure 2800.66 les Number: 97-041 CPS procedure 2800.66 is a new test procedure that will be used to verify that the Area Radiation / Process Radiation (AR/PR) System soRware changes made during the implementation ofPR.F018, Supplement I will correctly maintain AR/PR channel parameters. The purpose of this test is to verify replacement soAware e,orrectly communicates with the monitors. This is not the final acceptance test for modification PR F018. This testing will be performed during MODE

4. During this test, alarms actuated by any field unit disconnected from the Main Control Room (MCR) central control terminal (CCT) is considered to be invalid. During the performance of this

- procedure, Radiation Protection (RP) and Chemistry personnel will perform local monitoring --

functions as necessary to meet all AR/PR system operability requirements. Disabling the CCT communication link requires actions to be performed locally at the monitor per plant procedures.

The main action in this procedure is disconnecting the field monitor strings from the MCR CCT and connecting them to the test CCT in the Radiation Protection Office. Thl action affects the ability of the MCR CCT to communicate with the AR/PR field units. It does not impact the ability of the safety related field units to perform their safety function since these field units have a direct connection to MCR panels and do not require a CCT for operation or for performing their safety function. Therefore, this procedure would net increase the probability of equipment malfunction. The accidents in the USAR do not rely upon operator action based on AR/PR system alarm annunciations. As such, this would not affect any of the accidents described in the USAR. The safety-related portions of the AR/PR system are designed to operate with the MCR 3

CCT unavailable; therefore no new failures are expected and implementation of the test procedure would not result in any new accidents or malfunctions. Operation of the AR/PR field units to perform their aarety function la independent of the operability of the MCR CCT Since this test is being performed during MODE 4 and the field units and local monitoring will be used to meet AR/PR systein operability requirements, the margin of safety has not been reduced.

i

. e o

Attaciuneet 2 ,

10 U4028% l Pase 90 of 142  :

, REACTOR COOLANT SYSTEM LEAKAGE TE3T TEMPORARY MODIFICATION  !

i Document Evaluated: CPS Procedure 9059.01 Log Nomber: 97 043-  !

i

This is a change to CPS procedure 9059.01, " Reactor Coolant System Leakage Test." The  !

chary,e adds new steps to install temporaryjumpers on the IB21 F032A/B valves to pressurize to the next downstream _ valve thereby testing the entire IB21 F032A/B valve assembly for leakage.  !

The boundary for this test has been the next downstream valve (IB21 F065A/B) However, l

without the foodwater check valves jumpered open, the check valve located at the approximate  ;

center of the assembly would maintain reactor pressure. Neither the downstream side of the l IB21 F032A/B valves nor the IB21F065A/B valves would be exposed to reactor pressure.

Therefore, the procedure did not test the IB21-F032 feedwater check valves realistically as part  ;

of the original hydrostatic test. This changes the original intent of the previous procedure used to i

perform this task, CPS procedure 2800.03, " Reactor Coolant System Leakage Test." This test  !

. lineup is being used as a post maintenance test such that the feedwater check valves can be hydrostatically tested aAer maintenance. -

l l

This temporary change will not cause malfbnctions of design basis equipment since no piping is ,

exposed to pressures different than previously experienced. Since this test does not change the i boundaries maintained for the test (IB21-F065A/B (feedwater inlet shutoff), IE12-F053 A/B i (RHR shutdown cooling injection) and the Reactor Water Cleanup (RT) System valves), the i change would not increase the likelihood of design basis accidents. Since this does not change the test pressures or boundaries of the test, there would not be a potential for creating a new type of

equipment malfunction. The containment isolation function is still maintained by the 1B21 F065A/B valves. Since the containment barriers would not be affected, the margin of safety i

would be maintained.

l REPLACE MOTOR OPERATED VALVE (MOV) ISX063B l

l Docunent Evaluated: ECN 29741 Log Number:- 97 045

\

This ECN resulted in the replacement of the SMC-04 actuator (nearing end-of life cycles) on butterfly valve ISX-063B, the Division II heat exchanger discharge valve with a more efficient SMB-000 actuator. The ECN also changes the valve orientation with respect to the installed pipe to account for actuator interferences. The stroke time will increase as a result of this change as i the gear ratio changes from 67.2:1 to 100:1. The new actuator torque has been evaluated and l determined to be suitable for this valve. This will change the facility with respect to the heat load h of the diesel generators that is rejected to Shutdown Cooling Water (SX) System cooling water.

Flow testing performed on the valves determined that flow results were acceptable for the ISX- ,

063B valve. In addition, since there is a delay in the development and transfer of heat load from  !

. the diesels to the heat exchanges, when the diesels first starts, the increased stroke time would not adversely affect the function of the SX system regarding heat removal capability or characteristics.

- The seismic, pipe stress, voltage requirements and fire load impacts were reviewed. This change did not adversely affect any of the above mentioned qualifications.

I l

m .mm.,_,---.e,re, t. .a,,, , emw-,,-.r.~.ce-3.,_%, .-w, .

, , , _-~ .s-,, -.y._.__,._,-,-,-..,m-,...-..,.c.- .- -ee.- - - ---

4 Auaciumsat 2 to U402sM Page 91 of142 Due to the similarity of fbaction between actuators and review of qualification acceptability, this change would not impact design basis equipment malfunctions. Failure of the Diesel Generator i (DO) oxtemal cooling source is not an initiating design basis accident in the USAR. W change I would be bounded by the USAR accident which assumes the loss of one divisional power supply. ,

New accidents or equipment malfunctions would not be introduced by this change since the  !

change resuhs in a valve assembly qualified to the environmental conditions and is identical to  !

other safety related MOVs already installed at CPS. W fbnetion and operation of SX, DG and  !

interacting systems as described in the Technical Speel6 cation Bases will be unaffected by this -

change. Therefore, the margin of safety would not be reduced.

NUCLEAR PERFORMANCE CALCULATION Document Evaluated: USAR Section 7.7.1.7.5.1 Log Number: 97 046 This USAR section is being clarified so design document requirements are accurately identified.

- This changes subsection (6) from, "W LPRM amplifier gains are not to be physically altered  ;

except immediately prior to a whole core calibration using the traversing in-core probe (TIP)  !

system." to, "N LPRM amplifier gains are not to be adjusted except during the performance of a i calibration." Interpreted out of context, the mention of the TIP system implies that a specific order of performance is required when adjusting the gains. Gain adjustments have to be performed after the TIP system is run using the process computer OD 1 program. The OD-1 computer design specifications state, "It shall be operationally forbidden to tamper with the gain adjustments on the LPRM amplifiers, except immediately prior to a new whole-core calibration procedure." This sequence is required to obtain accurate values for the gain adjustment, The process of adjusting the local power range monitor (LPRM) involves increasing the sensitivity _ of the LPRM such that its reading matches the nonnalized TIP detector readings as the TIP passes the LPRM. The gain adjustments are available after the TIP system is run. Using the gain adjustment factor before the TIP system is mn would result in using TIP data as much as six weeks old.

A second change to this same subsection eliminated the words, "to the operator". This change is to maintain consistency with USAR Chapter 13," Conduct of Operations." W Chapter 7 operator is a computer operator unlike the Chapter 13, " Licensed Operator "

This change clarifies the USAR for conslatency with design documentation.' The LPRMs are calibrated in accordance with the design specifications and licensing documentation. This change does not affect the LPRM calibrations and would not affect design basis equipment malfunctions.

Nuclear performance calculations do not have design basis accident functions. This change eliminates the potential for misinterpretation of the functional requirements of nuclear j

performance calibration. With the LPRMs correctly calibrated, there is not a possibility for creating new accidents or equipment malfunctions. Making this clarification to the USAR has no ,

impact on the margin of safety for any technical specification. ,

3

Me% 2 to U4028%

Page 92 o(142 CHANGE IMPLEMENTATION REQUIREMENTS FOR TEMPORARY PROCEDURE CHANGES Document Evaluated: CPS Procedure 1005.06 Log Numbeir: 97 048 This change revises CPS administrative procedure 1005.06, " Conduct of Safety Reviews,"

Section 8.2.5, " Implementation." This change allows the completion of a safety evaluation screening for temporary changes to procedures made in accordance with USAR Section 13.5.1.2,

" Preparation of Procedures," and Operational Requirement Manual (ORM) 6.8.3, " Temporary Changes," to be completed at the earliest practicable time (and within 14 days). This change will allow temporary procedure changes which do not impact the intent of the procedure to be made in a timely manner This evaluation is being performed to evaluate the impact of timing regarding the completion of the 50.59 process.

This procedure change involves no changes to plant equipment, physical plant changes or changes to plant operating procedures. As a result, the change would not impact design basis accident probability, equipment malfunctions, create new failure types or decrease the margin of safety.

ANSI Standard N18.7 1976, " Administrative Controls and Quality Assurance for the Operational

( Phase of Nuclear Power Plants," provides guldsnce for procedure adherence in Section 5.2.2 and steps to allow temporary changes which clearly do not change procedure intent. The ORM already has provisions to allow temporary changes to be processed in this manner. NRC Inspection Manual Procedure 42700, " Plant Procedures," Section 3.5.c outlines the method by which the NRC allows temporary changes which parallel this change. As such, this change to CPS procedure 1005.06 does not conflict with ANSI, ORM or NRC guidance.

REACTOR RECIRCULATION (RR) SYSTEM OPERATING PROCEDURE Document Evaluated: CPS Procedure 3302.01 Log Number: 97-051 This change to CPS procedure 3302.01," Reactor Recirculation," revises the use of and step order for the IB33 F075 seal staging valve. Specifically, the requirement of shutting the 1B33-F075 valys only aRor the loop has been cooled to less than 250 'F is being deleted, Instead, it is being changed to shut the valve any time that reactor recirculation (RR) loop isolation is required.

Engineering evaluations and vendor concurrence recommended tint the isolation take place shortly after the RR pump is isolated instead of waiting until the idle loop cools to less than 250

- 'F. The technical and safety advantages are: (1) more seal irsection flow will be forced down through the pump thereby minimizing the amount of high temperature real leakage, (2) it will allow the seal to withstand a higher system pressure before de-staging, (3) it will minimize the potential of future seal damage / leakage, and (4) allow for full cooldown of an idle RR loop, in addition, NUREG/CR-4544, " Reactor Coolant Pump Seal Related Instrumentation and Operator

~ Response, an Evaluation of Adequacy to Anticipate Potential Seal Failures," and General Electric Design document GEK K2801-0005 provide recommendations for closing the seal staging valve when the pump is shutdown, seal injection flow is lost, and component cooling water (CC) is lost.

Auschment 2 l la U 602836 j Page 93 of142 This change is intended to minimize the possibility of catastrophic loss of seal or seal de-staging.

As such, this change would not increase design basis equipment malfunctions. USAR Section 5.4.1.3 states that, "The pump shaR passes through a breakdown bushing in the pump casing to reduce leakage in the event of a gross failure of both shaA seals." This change does not impact the probability of this event. Since gross failure of the seals is postulated in USAR Section 5.4.1.3, this change would not create new accidents or equipment malfunctions. This change does not decrease any margin of safety since the IB33 F075 valve is not part of the analyzed reactor wola.at boundary within the USAR Section 15.6.5 LOCA design basis event.

INSTALLATION OF WS THROTTLING VALVES AND DELETION OF CHECK VALVE Document Evaluated: ECN 29055 LogNumber: 97-032 This ECN resulted in a throttling valve IWS253B being installed in the upper seal supply line and 1WS254B in the lower seal supply line of the IWS0lPB pump. These throttling valves were installed downstream of existing flow indicators IFIWS122 and IFIWS134 to control flow in the respective lines. A thiottling valve in each line will better control flow compared to the single .

valve present practice. This ECN also deletes existing check valve IWS085B which is not used to protect any equipment. This change rec ires revi6 ion to USAR Figure 9.2-1, Sheet 4, due to the addition of valves IWS253B and IWS..s4B and deletion of check valve IWS085B.

The Plant Service Water (WS) System is non safety and non seismic. Changes made are to the non-safety, non-seismic portion within the screenhouw. As such, adding throttling valves to better control flow and deleting an unnecessary check valve would not impact design basis equipment malfunctions. Since this change does not impact any safety-related systems, the change would not impact design basis accidents. This ECN does not affect the design flows and pressures to the pump seals. By monitoring the aforementioned flow indicators, this change allows the operation of the WS pump to be controlled within its design pararr.eters. 'I oerefore, this change would not have the potential to create a new type of accident or equipment

- malfunction. This change has no impact on the intended function of the WS system or the basis for any Technical Specification. Therefore, this change would not impact margin of safety.

INSTALLATION OF WS THROTTLING VALVES AND DELETION OF CHECK VALVE Document Evaluated: ECN 29056 LogNumber: 97-053

- This ECN results in throttling valves being installed (lWS253C) in the upper seal supply line and (IWS254C) in the lower seal supply line of the IWS0lPC pump, These throttling valves will be installed downstream of existing flow indicators IFIWS123 and IFIWS132 to control flow in the respective lines. Adding a throttling valve in each line will better control flow compared to the single valve previously used. This ECN also deletes existing check valve IWS085C which is not used to protect any equipment. This change requires revision to USAR Figure 9.2 1, Sheet 4, due to the addition of valves IWS253C and IWS254C and deletion of check valve IWS085C, The Plant Service Water (WS) System is non-safety and non seismic. The changes made are to the non-safety, non-seismic portion within the screenhouse. As such, adding throttling valves to

Attachment 2 -

O U 602 36 Page 94 of 142 better control flow and deleting an unnecessary check valve would not impact design basis equipment malkncticns. Since this change does not impact any safety related systems, the change would not impact design basis accidents. This ECN does not affect the design flows and

- pressures to the pump seals. By monitoring the aforementioned flow indicators, this change allows the operation of the WS pump to be controlled within its design parameters. Therefore, this change would not have the potential to create a new type of accident or equipment malfbnction. This change has no impact on the intended funct!on of the WS system or the basis for any Technical Specification. Therefore, this change would not impact margin of safdy.

OPERATIONS CONTINUING TRAINING PROGRAM DESCRIPTION REVISION Document Evaluated: Appendix D Revision Log Number: 97-055 This change incorporated commitment tracking numbsrs into Appendix D, " Program Commitments," of ti'e Operations Continuing Training Program Description. This does not require a USAR, Operational Requirements Manual (ORM) or Technical Specification change.

This is an administrative change to the Operations Continuing Training Program Description.

Changes of this nature, adding commitment numbers to an appendix, would not change design basis accidents, malfunctions or margin of safety.

REVISE OFFGAS (OG) FLOW RANGES PER CR 1 97-02-181 Document Evaluated: USAR Section 7.7.1.10.3.11 Log Number: 97 056 USAR Section 7.7,1,1b.3.11, "Offgas System Flow Measurements," lists the normal offgas flow range as 6 to 30 standard cubic feet per minute (SCFM) and the startup flow range as 30 to 300 SCFM. The as-built equipment has a flow range greater than the USAR listed range for both normal and startup conditions. This changes the USAR to correct the low end number for each of the offgas ranges. This :.hanges the normal range from "6 to 30 SCFM" to "3 to 30 SCFM" and the startup range from "30 to 300 SCFM" to "3 to 300 SCFM.". There is no change to the design limits of the offgas flows or any other equipment. No setpoints are being changed nor are there any changes to mai ual or automatic functions. There are no changes to the instrument: as operators will use the same flow scales.

Since this change does not impact equipment operation and offgas flow, design basis equipment

. malfunctions are not affected. USAR Section 7,7.1,10A 1 states that the gaseous radwaste control system is not required for safety purposes or after design basis accidents. Therefore, this change would not affect design basis accidents or frequency of them. The USAR Offgas flow range change does not change the system operating ranges or setpoints. As such, this change would not creat new design equipment malfunctions. The Offsite Dose Calculation Manual (ODCM) requira. ents are not being changed. Therefore, offsite dose requirements and margin of safety are unchanged.

Anachraent 2 to U 602836

, Page 95 of142 TEMPORARY MODIFICATION (TM) TO REMOVE AP-028 REGULATING TRANSFORMERS IN DIVISION 2 Document Evaluated: TM 97-017 Log Number: 97-057 Regulating Transformers, OP55EBRT,057ERT, I AP75ERT, a ,d 1 AP94ERT (installed by ModiAcation AP-028) associated with Motor Control Center (MCC) oAP55EB, OAP57E, I AP75E, and I AP94E respectively have experienced spurious tripping problems that are influenced by Electro-Magnetic Interference (EMI) and Radio Frequency Interference (RFI)

- noise. This temporary modincation provides attemate power to the circuits fed from tlese regulating transformers and de-energizes and isolates the regulating transformers. -The alternate power is provided by r connecting the non regulating transformers that were previously abandoned in place by AP-028.

- USAR Section 15.2.6 addresses Loss of Oss :e Power events. USAR Chapters 6 and 15 evaluate loss of power and loss of one diesel generator in corQunction with these accidents. The non.

regulating transformers were part of the original licensing basis under which CPS was licensed.

Therefore, tiv, re-installation of these ncoregulating transformers does not resuk in the probability or consequences of events involving the lou of power previously evaluated.

AP-028 installed the regulating transformers, to replace the non regulating transformers, and to provide voltage support in case of a degraded voltage condition. The non-regulating transformers were part of the original design basis for CPS and were purchased Class 1E. There have been no changes to the non regulating transformers, since being abandoned in place, other than having their terminals disconnected and removed.

- Since there have been no changes to the non-regulating transformers to affect their safety or environmental qualification, the probability or consequences of failure of equipment important to safety is no different than that cf the original license. Similarly, since there have been no changes to the transformers since being abandoned in place, the possibility of a new type of accident is no greater than that of the original license.

USAR Section 8.3.1 discusses degraded voltage conditions and how a degraded voltage condition would be mitigated by degraded voltage protection relays that would isolate IE busses from a degraded offsite source. Technical Specification 3.3.8;l provides the required actions when these relays are inope able. Technical Specification 3.8.10 provides the required actions when the bus voltage is degraded. Therefore, replacement of the regulating transformers with the non.

regulating transformers, while operating under the Technical Specifications ensures that the margin of safety is not reduced.

t INSTALL TEMPORARY PLUGS ON THE AUXILIARY BOILERS A & B Document Evaluated

Temporary Modification (TM)97-021 LogNumber: 97-058 This TM is to facilitate the r'epair s f chemical feed pump OAS06PB located on the auxiliary boiler

skid. The TM installs temporary caps and/oi plugs on the suction and discharge piping of the

Attachment 2 to U 602836

Page 96 of142

. pump when it is removed for maintenance. Thl TM is needed as the discharge of both pumps is cron-connected without manual isolations. CPS procedures and dealgn only permits one "on-j line" chemical feed pump at a time; the other pump must be "off line." Once maintenance is complete, this TM will be removed. This is a change to the system configuration as described in

' US AR Figures 3.6 1 and 9.5-6 in addition to the description in USAR Section 9.5.9.2. i USAR Section 9.5.9.3 states, "W auxiliary steam system has no nuclear safety-related function". <

Therefore, this change would not increase the probability or consequences of design basis l 4

equipment malfhnetions. W caps / plugs are installel in a low energy portion of the auxiliary steam system as defined by USAR Section 3.6, associated with dynamic effects of high and

. moderate-energy line ruptures, and Figure 3.61, Sheet 30. Therefore, this modification would  ;

not impact high and moderate rupture analysis. If any malfunction would occur with one pump in operation, the auxiliary steam boilers would be shut down per normal plant procedures with no impact on the safe operation of CPS. As such, this would not create the possibility of a new

[ accident or malfunction of equipment. % Auxiliary Steam System is classified as non-safety; i therefore, removal of the "B" chemical feed pump and installation of caps / plugs would not impact ,

[ the margin of safety.

REPAIR HOLLOW BLOCK WALL TO A QUALIFIED 1.9 HOUR FIRE RATING Document Evaluated: ECN 30131 LogNumber: 97-059

This ECN repairs a masonry block wall in the Control Building. This repair will restore the wall

to a qualified 1.9 hour1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> fire rating. The wall currendy is 7 - 5/8-inch hollow block construction but has = 5-inch diameter penetration on one side to accommodate an interfering pipe hanger. This ,

change replaces two of the hollow blocks with four solid masonry blocks, notched to accommodate the pipe hanger. However, this will provide greater than 4.1 inches of block

- equivalent to the wall. The wall at this locatien acts as a fire wall to separate Division Ill i

electrical equipment from cable risers associated with Divisions I and II. USAR Appendix E,

. Sections 3.4.5.1 and 3.4.5.3 identifies these walls as " hollow," concrete, The USAR is being revised to reflect the change made by this ECN. ,

%e additional mass of concrete is negligible and is bounded by the current structural calculations.

l - Therefore, this change will not increase the probability or consequences of dedgn basis equipment

{ malfunctions. The change will decrease the probability or consequence of design basis accidents .

4

. as this change fully qualifies the fu rating to this wall. Since the wali a structurally qualified, the change would not produce any credible new equipment malfunctions. Fully qualifying this wall to

a 1.9 hour1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> fire barrier will not reduce any ma4 gin of safety, i

6900/4160/480V SWITCHGEAR/ CIRCUIT BREAKER OPERABILITY PROGRAM Document Evaluated: CPS Procedure 1014.11, Rev. O Log Nu.nber: 97-060 2

CPS procedure 1014.11, "6900/4160/480V Switchgear/ Circuit Breaker Operability Prograin,"

, provides operating guidance for the operability of breakers when safety-related switchgear breakers are in a seismically unanalyzed configuration. The racking out and removal of breakers results in a -

i-

Ata=A ===42 is U402sw Page 97 of142 change from the normal operating and seismically qualified configuration of the switchgear. The normal confistiration is when the breakers are racked in and the cubicle doors are closed and fastened. During performance of on line surveillance and maintenance activities the potential exists for the affected switchgear to be in a condition that was not previously seismically qualified to meet USAR Section 3.10 requirements. These requirements are to meet IEEE Standard 344-75, "Recommemled Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations," and Regulatory Guide 1.100, " Seismic Qualificatiore of Electric Equipment for Nuclear Power Plants." This guidance is needed since documentation of seismic type-testing or analysis is not available for the following conditions: (1) When breakers are racked-out, racked-

-down and left in the cubicle; (2) Removed from the cubicle; or (3) When the cubicle doors are opened during normal operation of the switchgear. This procedure addresses safety related Division I and II, 4160V, 6900V switchgear, Division III High Pressure Core Spray switchgear, and 480V unit substation switchgear in the unit auxiliary power systems.- This does not apply to non-safety related switchgear in seismic category I builoings.

This procedure will define the conditions under which the safety-relt.'.ed switchgear is seismically qualified. The procedure will ,.. "' administrative control over entry into seismically unanalyzed conditions (only permitteJ when redundant equipment is operable and personnel remain by the breaker cubicle (s) for which the condition exists). The procedure will also establish a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> time limit for when these seismically unanalyzed conditions may exist before declaring the associated switchgear inoperable; and provide operational guidelines and instructions to track the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> limit. The Division I and 11480V AAB unit substations are seismically qualified with the breaker in the racked-out and test position, and left in the cubicle (supported by lilinois Power calculation IP-Q-0391 R/0). For Division I and 114160V/6900V switchgear, 25% of the breakers are permi,tted to be removed for a limited time (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />). For Division 111,4160V switchgear, no more than one breaker is allowed to be removed for a limited _ time (48-hours).

USAR Section 8.3.1 describes the Class IE power systems. However, the USAR does not specifically address breaker position. The risk of having switchgear in a seismically unanalyzed condition is small due to the risk of earthquake and limited time which the equipment is in this configuration. The improvement in reliability and availability expected to be gained by performance of maintenance would enhance sdety more than the limited risk as mentioned above.

Therefore, this procedure would not impact design basis equipment malfunctions. A oeismic event is an accident evaluated in the USAR. However, the USAR does not address the impact of a seismic event to qualification of switchgear during maintenance activities. This on-line maintename should increase electrical system reliability. As such, this would improve the likelihood for mitigation of design basis accidents.

USAR Chapter 15A does address allowable repair time rules for failed equipment. However, this section does not specifically address seismically unanalyzed conditions. Limited seismic testing demonstrated that the switchgear's intended function was not compromised when the assemblies were subjected to earthquake intensities far greater than the CPS design basis earthquake. This supports the fact that allowing these configurations for short durations would not introduce the potential for new equipment malfunctions. Performing on-line maintenance would not by itself create a gwtontial for a new accident. Core damage probability and margin of safety would not be affected due to: the low frequency of earthquake, short duration of condition (48-hours), multiple

. .-_-1

Attachunent 2 to U402836 Page 98 of142 redundant equipment failures /unavailabilities would have to occur concurrently to make Emergency Core Cooling System (ECCS) functions unavailable, and no degradation of operating switchgear after an earthquake.

USAR CHANGE FOR NSPI DEPARTMENT AND ADDITION OF THE ASSISTANT VICE PRESIDENT Document EvalustM: USAR Chapter 13 LogNumber: 97-061 This USAR change is to reflect organizational changes at CPS due to the formation of the Nuclear Safety and Performance improvement (NSP1) Department and the addition of the Assistant Vice President (VP) position. This change add 6 responsibilities of the Assistant VP and the Manager NSP1 to Section 13.1.1.2. The Human Performance Enhancement Coordinator and Corrective Action Program Administrator functions were removed f om Section 13.1.2.2.5. The Assistant VP and Manager NSPI were added to Figure 13.1-1 A and the Assistent VP was added to Figure 13.13.

These changes consolidate assessment groups at CPS. The NSPI will also include a group

- dedicated to develop, coordinate and monitor site wide improvements. The movement of Corrective Action Program from Plant Staff to the NSPI Department is in compliance with both ANS 3.2, ANSI N18.71976,"Administintive Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants," and ANSI /ASME N45.2 1977, " Quality Assurance Programa Requiremeats for Nuclear Facilities." The addition of the Assistant VP position does not advertently impact the duties and responsibilities of the VP-Nuclear as defined in the Technical Specifications, Operational Requirement: Manual (ORM) or US AR.

These changes are administrative in nature and are not directly related to equipment manipulation.

Therefore, this change would not impact design basis equipment malfunctions. The functions and activities of the groups comprising the NSPI Department continue as before. At such, the change would not involve new accidents or equipment malfunctions. This change does not reduce the margin of safety.

WS TO CC DIFFERENTIAL PRESSURE AT THE CC TO WS HEAT EXCHANGERS (HXs)

Document Evaluated: USAR Section 1.2.2.8.3 Log Number: 97 062 USAR Section 1.2.2.8.3 states that the Plant Service Water (WS) system operates at a pressure higher than the Component Cooling Water (CC) system to prevent potentially radioactive water from entering WS. In addition, it states that a radiation monitor is provided to monitor the discharge. The design criteria for the WS system have similar statements concerning WS pressure being higher than CC. Under most plant operating conditions, WS pressure is not always higher than CC pressure at the interfacing heat exchangers (HXs). It was determined by condition report 1 96-10-260 that revision to operating procedures would not correct the problem due to equipment limitations. The US AR is being revised to delete the statement of" higher pressure."

Anaciunent 2

. $ 602836 Page 99 of142

UtiAR Section 9.2.1.1 states that WS is not required to ensure safe shutdown. USAR section

. 9.2.2.3 states that CC is not required for safe shutdown. His change would not impact )

probability of accidents; this change is primarily concerned with increased consequences. W l plant does have the capability to manually isolate a leaking WS to CC HX using installed valves.  !

j The closed loop CC system operating in conjunction with the existing radiation monitorinc i 1

_ instrumentation provides adequate capability to prevent unacceptable releases to the environment.

To ensure adequate response to monitor alarms and failures, Clinton Power Station is adding the i CC process radiation monitor (PR037) into the Offsite Dose Calculation Manual (ODCM). The

. ODCM already states in 2.3.5 that releases from the WS are considered an abnormal event and will be accounted for in the, " Radioactive EfHuont Release Report." Therefore, this change would not increase the consequences of an accidert or malAar.ction cf design basis equipment.  ;

Only the interfhce at the Fuel Pool Cooling and Cleanup (FC) HX between CC at.d SX is safety.  !

- related. This change does not affect this interface. As such, this change would not create new  ;

- design basis equipmer t malfhnetions or accidents. The Technical Specification (TS) Bases only  ;

require these systems be capable of remote operation from ti.e Maln Control Room. This has not .

changed. The change would not impact 10 CFR Part 20 limits or the release limits in 10 CFR

. Part 50, Appendix L Therefore, thh change would not decrease the margir. of safety.

) REPLACEMENT OF DIVISION 1 SECOND LEVEL UNDERVOLTAGE RELAY AND

TRANSFORMER TAP SETTING CHANGES '

Document Evaluated

Modification AP-027 Log Number: 97 064

This modification will replace the existing second level under voltage relays with more accurate L ones. It will also change the tap settings on potential transformer lHG03JA to ensure sufHelent l voltage is available for the hydrogen recombiners. The modification also makes two minor i

changes, the first is the replacement of the alternating current (AC) elapsed time meters with direct current (DC) elapsed time meters; the second minor change is the addition of a relay test l

switch. The goal of this modincation is to assure that under LOCA loading conditions, there is t sufHcient voltage at all the Engineered Safety Feature (ESF) equipment such that it will start and ,

run. This modiAcation will bring CPS into compliance with the requirements of 10CFR50 ,

3 Appendix A, GDC 17 and assure that there is adequate voltage from the off site power system.

The modification meets the requiremems of the Standard Review Plan and the NRC Branch h Technical Position PSB 1 contained in NUREG 0800, " Standard Review Plan for the Review of '

Safety Analysis Reports for Nuclear Power Plants." This modification is consistent with the CPS ,

i response and commitments made in the IP response ta USAR Q&R 430.135 and documented in '

l the NRC Safety Evaluation Repon (SER) NUREG-0853, " Safety Evaluation Report Related to l the Oporation of Clinton Power Station, Unit 1," Supplement 7. The degraded voltage issue was i recog dzed and reported in LER 94-005 date June 3,1994, if - By letter dated December 4,1996, NRC approved Amendment 110 to the CPS Facility Operating l License. This modiScation implements the revised degraded voltage setpoints of Amendment 110 ,

j for Division 1, Plant modifications AP-028 and AP 029 implemented the setpoints for Division 2

and 3, respectively. Modifications AP-028 and AP-029, corresponding to Safety Evaluation Log Numbers96-015 and 97 069, respectively, a~e included in thh report. All three modifications are similar. However, the Division 2 modification, used regulating transformers and abandoned the

.,, n _.N a.

.m.~ . N- - . 4. n , , . _ , ,-._._,.,,n w.4 4 ,. .,-,..m...,.,.. ,., , ,, ., ,,ge..,,., ww. . .e n a n y w.m ., r .--,.,n--. y,-,.m.,,,.mpp-_ -r.-v.p,

Aw% 2 O U402s%

Page 100 of142 existing non-regulating transformers in piace. A$er lutallation, the regulating transformers were fou:ul susceptible to radio firoquency and electro-magnetic interference. ModiAcation AP 027 uses b existing non regulating transformers.

The equipment uwd in this modiAcation will not increase the probability of a malfunction of equipment imponant to safety. The new relays are qualified Class lE por IEEE Standards 323 1974, " Standard for Qualif'ying Class IE Equipment for Nuclear Power Generating Stations," and 344 1975, " Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations." The failure modes ase the same as the old equipment. The ,

consequences of a malfunction are the same as previously analyzed because the failure modes are the same as analyzed in the odginal design. Since this modidcation implements an approved change to the technical specifications, which has been reviewed, the change does not reduce the margin of safety.

CONSOLIDATION OF OPERATING EXPERIENCE (OE) RESPONSIBILITIES Document Evaluated: USAR Appendix D,I.C.5 LogNumber: 97-065 This revises the USAR to consolidate b responsibility for coordinating the review of Operating Experience (OE) Information to the Licensing Department. This change combined the function of

CPS procedure 1006.04, " Review of Operating Experience," canceled in December 1995 into l Licensing Procedure L.1, " Feedback Program." Procedure L.1 addrepes the program for the review of OE information. Whereas CPS procedure 1006.04 had sa Assistant Power Plant i Manger " direct" the prog am, L 1_ assigns that responsibility to department managers. Procedure

L.1 also directs that only " pertinent, non-conflicting informaticn," be disseminated; this is a more

specific sorting and assignment of OE reviews. This change would not affect equipment j operation; therefore, design basis equipment malfunctions would not be changed. This is a change j to an administrative procedure. As such, accidents evaluated in the USAR would not be affected j by the change. Since procedure L.1 does not directly impact the operation, maintenance, or 1 control of CPS structures, systems, or components, this would not create new equipment

! , malfunctions or accidents. This does not impact the margin of safety as the feedback program does not perfonn any functions described in the basis for any Technical Specification l

REVISION TO USAR AND TS BASCS ECCS RESPONSE TIME STATEMENTS l Document Evaluated

USAR Section 9.2.1.2.2, TS Bases 3.5.1 - LogNumber: 97-066 The Technical SpeciAcation (TS) Bases are being n evisal to replace references to USAR Table l 6.3-8 with the specific emergency core cooling system (ECCS) subsystem response time limits.

- This is being done to eliminate potential confbslon over the appropriate limits. The USAR is being revised to eliminate potentially confusing response time statements and correct references to USAR Tables containing ECCS response times. In addition, the USAR is being revised to 1

eliminate reference to periodic measurement of" pump run-up" times. The TS Bases are being i revised back to the format used prior to implementation ofimproved technical specifications by

[ specifying the overall system time limit previously established by safety evaluation Log Number L

l u

Attaclument 2 to U4028%

Page 101 of142 94-0057, reported in Revision 6 to the USAR. These changes are being made in response to NRC Notice of Violation 50-461/96015 05b.

j Pump run-up times are considered analytical values and not performance criteria to be measured.

Calculations confirm pump run up times to be leu than 5 seconds even under degraded voltage conditions. The ECCS lidection valve stroke time limits are less than or equal to 15,19 and 20 seconds for High Pressure Core Spray (HPCS), Low Pressure Core Spray (LPCS), and 14w Pressure Coolant Irdection (LPCI) subsystems, respectively. ECCS response time testing is required to be performed every la months on each subsystem. Other testing such as quarterly inservice testing provides adequate assurance it.st the pump run up times remain well within the stroke times measured for the ledection valves. As a result, actual ECCS response times will not be impacted by this change nor would the probability or consequences of design basis equipment malfunctions.

USAR Sections 6.2 and 6.3 address ECCS performance in mitigating LOCAs. Since the ECCS response times would not be impacted by this change, neither would accidents evaluated in these USAR Sections. This change does not involve a change to the plant equipment or its operations.

Therefore, this change would not create new equipment malfunctions. USAR Section 15.5.1 evaluated an inadvertent HPCS pump start; however, as stated in US AR Section 15.5.1.3.3.1, "no delays were considered because they are not relevant to the analysis." Neverthelen, ECCS response time, run-up time and stroke times will still be met through required testing. These changes will ensure that the megin of safety has not been reduced.

DELETE STATEMENT THAT SX OPERATES AT A HIGHER PRESSURE THAN FC Document Evaluated: USAR Appendix D, I.C 5 Log Number: 97-067 USAR Section 9.2.1.2.2 states that the Shutdown Service Water (SX) System operates at a pressure higher than the liuel Pool Cooling and Cleanup (FC) System to prevent potentially radioactive water entering SX Rom FC. Review of calculations and other design documentation determined that the SX pressure (approximately 97 psig) would be lower than the FC pressure

- (approximately 120 psig). However, it has been determined that operation of the two systems in a manner presently descrioed in the USAR is not possible. Therefore, "SX operates at a higher pressure than FC," is being deleted Rom the USAR.

Because the diff'erential pressure (between the SX and FC systems) is small, the probability of creating a leak has not increased. Therefore, the change does not increase the probability of equipment malfuncticn Process radiation monitors are located in the SX discharge from the Residual Heat Removal (RHR) heat exchangers (HXs). If a high radiation alarm occurs, the operator can isolate the HX by closing the SX water inlet and discharge motor-operated isolation

< valves. The second RHR or FC HX could then be placed into operation. This monitor is a part of the process radiation monitoring system described in USAR Section 11.5. This will ensure that release limits are not exceeded. As such, consequences of a design basis equipment malfunction would not be changed.: USAR Sections 9.2.1.2.2 and 11.5.2.2.8 consider FC to SX HX leakage as an anticipated operat'ng occurrence for which detection and isolation capabilities have been provided. A leak from FC to SX is the only equipment malfunction which could occur from this

Attachaient 2 to U4028%

Page 102 of142 change. Since this leak is already considered in the USAR, this change would not create a new type of equipn.ent malfunction. Technical Specification (TS) Sections 5.5.1, "Off' site Dose Calculation Manual (ODCM)," and 5.5.4, " Radioactive Effluent Controls Program," were reviewed. No references or statements were found that indicate the TS or TS Bases relied upon maintaining SX pressure higher than FC and the FC HXa, Therefore, the musin of safety is not reduced by this change.

REPLACEMENT OF DIVISION 111 SECOND LEVEL UNDERVOLTAGE RELAY AND DISTRIBUTION POWER SUPPLY Document Evaluated Modification AP-029 - Log Number: 97-069 This modification will replace the existing second level imder voltage relsys with more accurate relays. The goal of this modification is to assure that under LOCA loading conditions, there is sufficient voltage at all the Engineerei Safety Feature (ESF) equipment such that it will start and run. This modification will bring CPS into cc'npliance with the requirements of 10CFR50 Appendix A, GDC 17 and assure that there is adequate voltage from the off-site power rystem.

The modification meets the requirements of the Standard Review Plan and the NRC Branch Technical Position PSB-1 contained in NUREG 0800, " Standard Review Plan for the Review of S Jety Anwysis Repcrts for Nuclear Power Plants." This modificatio . is consistent with the CPS response and commitments made in the Illinois Power (IP) response to USAR Q&R 430.135 and documented in the NRC Safety Evaluation Report (SER) NUREG-0853, " Safety Evaluation Report Related to the Operation of Clinton Power Station, Unit No.1," Supplement 7. The degraded voltage issue was tscognized and reported in LER 94-005 date June 3,1994 By letter dated December 4,15%, NRC approved Amendment 110 to CPS Facility Operating License. This modification implements the reded degraded voltage setpoints of amendment 110 for Division 3. Plant modifications AP-028 and AP-029 implemented the setpoints for Division 2 and 3, respectively, Modifications AP-028 and AP-027, corresponding to Safety Evaluation Log Numbers96-015 and 97-064, respectively are included in this report. All three modifications are similar. However, P.a %ision 7 modification, used regulating transformers and abandoned the esisting non-regulating transformers in place. Anet inst.Jiation, the regulating transformers were ,

found susceptible to radio frequency and electrsmagneuc interfwence. Modification AP-029 uses the existing nn-regulating transformers.

The equipment used in this modification will not increase the probability of a malfunction of equipment important to safety. The new relays are qualified Class IE per IEEE Standards 323-1974, " Standard for Qualifying Class IE Equipment for Nuclear Power Genecation Stations,"

and 344-1975, " Recommended Prg.cti:es for Seismic Qualification of Class IE Equipment fer Nuclear Power Generating Stations." The failure modes are the same as the old eqinipment. The consequences of a malfunction are the same as previously analyzed because the failure modes are the same as anDzed in the original design. Since this modification implements' an approved change to the tec' nical specifications, v hich has been reviewed, the change does not reduce the margin of safety.

l l

Attachmosa 2 to U.402836 l Page 103 of 142 REPLACE CYCLED CONDENSATE (CY) SYSTEM RELIEF VALVE (GL 96-06 ACTION)

Document Evaluated: ECN 30156 Log Number: 97-070  ;

ECN 29913 resulted in the installation of a relief valve ICY 124 on the Cycled Condensate (CY)

System water header ICY 28C-6 in containment. The relief valve relicf presrm ms 140 psig.

The relief valve was installed in response to Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions," concerns on overpreuurization of containment penetrations due to thermal expansion following a post-design t basis LOCA heatup. W installation of the relief valve was evaluated by Safety Evaluation Log -

Number SE 96-076 and reported earlier in this report. ECN 30156 replaces the 140 psig relief valve with a 190 psig relief valve. ARet installing the 140 psig relief valve, the CY header (ICY 28C-6) pressure was found to be 130 psig. W relief valve was replaced to minimize valve cycling due to normal pressure fluctuations and the closeness of the setpoint of the previous valve to the header pressure. W 190 psig valve is installed in the same location, will relieve to the same place, and is constructed of the same material as the previous valve. Therefore, the analysis of SE %-076 remains valid. ,

USAR CHANGE FOR SUPPRESSION POOL MAKEUP (SM) SYSTEM Document Evaluated: USAR Section 6.2.7.3.1 LogNumber: 97 071 USAR Section 6.2.7.3.1 has been revised to indicate that automatic actuation of the Suppression Pool Makeup (SM) System occurs when suppression pool level is between 0.8 and 18 inches .

- below normal low pool level. This change was made to account for instrument driA, loop and [

instrument ac,:uracy. A delay time of 7.4 seconds has been calculated between the start of f Emergency Core Cooling System (ECCS) flow and dumping of the upper pool to assure drywell vent submergence, i- This change will not impact the probability of design basis equipment malfunctions since it does not change the method with which equipment performs its functions. USAR Table 6.2-5 l sui

• prizes that peak drywell pressure occurs at 1.45 seconds for a recirculation line break and 1.L. monds for a main steam line break, h time delay is still suHicient to ensure that the peak presmre has elapsed as described in USAR Section 6.2.7.3.1. Therefore, this change does not .

impact design basis accidents. Reducing the time from 2.75 minutes to 7,4 seconds would not create new accidents or malfunctions since USAR Section 6.2.7.3.3, " Inadvertent Dump," (no delay) concurrent with a LOCA bounds this change. Technical Specification (TS) Bases 3.6.'2.4 discusses the basis for the setpoint. This change makes the USAR consistent with the TS Bases; therefore, this change would not reduce the margin of safety for any Technical Specification.

i REPLACE MICROWAVE SYSTEM WITH FIBER OPTIC SYSTEM

! Document Evaluated: Modification SY-012 LogNumber: 97-072 i

Illinois Power (IP) is installing a second 345 KV transformer at the Brokaw Substation for IP's transmission rystem reliability and capacity. In addition, the existing microwave system for the

Niaciuneet 2 to U402sw Pase 104 of142 protection of the Brokaw 345 KV transmission line is being replaced with a fiber optic system at ,

the Brokaw Substation. Modification SY-012 evaluates replacement of the existing microwave '

system with a fiber optic system. This modincation will enhance the capability of the Brokaw line as a CPS offsite power systan. This change revises USAR Sections 1.2.2.8.17 and 9.5.2.2.5 to describe this new fiber optic system far the protective relaying of the transmission system.

USAR Sections 8.1 and 8.2 describe the offsite power transmission system and state that they comply with NRC's Gee. oral Design Criteria (GDC) 17, " Electric Power Systams," and GDC 1d,

" Inspection and Testing of Electric Power Systems." [This modification does not change compliance to these GDCs or conclusions in the transient stability study. However, the stability study for this modi 6 cation impads existing figures in USAR Section 8.2 which are being revited -

to reflect this modification.)

The function of the protective relaying system is unchanged. Only the method of transmitting the signals for the relays to activate has changed. The fiber optic system is more reliable than the micrc vave system, therefore, this change would not increase the consequences of equipment malfhnetion or accident. Since the electrical system still complies with GDCs 17 and 18, the -

change would not increase the probability of equipment malfunction or accident. The failure of the fiber optic system would be equivalent to a failure of the present microwave system; therefore, this would not create a new equipment failure or accident type. This change will enhance the reliability of the protective relaying and not impact USAR Section 15.2.6, " Loss of AC Power."

Therefore, margin of safety as defined in the Technical Specification Bases wouhi not be reduced.

REMOVAL OF TWO DRAIN LINES FROM THE WS SYSTEM Document Evaluated: MWR D62956 Log Number: 97-074

- Butterfly valves 1WS109 and 2WS109, WO chiller bypass valves and associated downstream piping were replaced during RF-6. Two 3/4 inch drain lines (lWS69A and 2WS698) attached to these lines were eliminated as directed by a field configuration change for MWR D62956. USAR Figure 9.2-1 is being revised to reflect this change.

The Plant Service Water (WS) System is a non-safety, non seismic system not required for safe shutdown of the plant This does not affect operation of WS as this piping can still be drained during plant outages by using plant Chilled Water (WO) System chiller tube side drains, There is no need to drain this WS piping during normal plant' operation. Eliminating these two drain lines does not affect the operation of the WS system. Therefore, previously evaluated equipment

. malfunctions or accidents would not be affected by this change. Since this change does not have impact on safety-r9ated stmetures, systems, or component _s, there would be no reduction in any _

margin of safety.

DESCRIPTION / DEFINITION FOR FLOW MEASUREMENT (CFM/SCFM/ACFM)

Document Evaluated: USAR/ORM/TS Bases Log Number: 97-075

- This change addej defmitions to help alleviate confusion when readings stated in cubic feet per minute (CFM) n ny be converted to Standard CFM (SCFM), Actual CFM (ACFM) or when no N

I -- - - - - -

f Attachment 2 OU402836 Page 10$ of142 correction is required. This adds a deAnition of CFM to USAR Appwlix A, " Glossary Table of Contents," revises Operational Requirement: Manual (ORM) 2.4.3, " Control Room Ventilation Smoke Mode, Testing Requirements," Section 4.4.3; and clarifies Technical Specification (TS)

Bases 3.6.4.3, Standby Oss Treatment System, and 3.7.3, Control Room Ventilation System.

This change will not increase the probaulity of a malfunction of design basis equipment because the change clarifles the testing of HVAC systems and does not alter equipment operation. The change provides guidance on testing to ensure systems operate per the original design. Therefore, this change would not change consequences of equipment failures or design basis accidents. u Since misoperation of these systems would not cause a design basis event, this change would not create 4 di% rent malfunction or accident. This clarification will ensure surveillance girements are ratisfactorilv perfonned. Therefore, tnis change would not reduce margin of safety.

FREEZE SEALS TO ALLON REPLACEMENT OF VALVES ISX014A & ISX014B Document Evaluated: MWRs D73305 & D733007. OPS 97 001 LogNumber: 97-076 The Plant Service Water (WS) and Shutdown Service Water (SX) Systems will be affected by the installation of a freeze seal. The freeze seal will serve as a temporary block to isolate the WS from the inoperable Division 1 or 11 SX system. Ti'e WS system will continue to supply all station requirements that are normally in service, except for the isolated division of SX. Using a freeze seal to isolste the WS from the SX systems la considered to be a short term change to USA".

Figure 9.2 2.

This freeze seal would not affect design basis equipment malfunctions since only one division of SX is allowed out of service at a time and Operations Coordination Plan OPS-97 001 coordinates contingency actions to mitigate a potential flooding event. During this evolution, water tight integrity of the operable division of SX is required. This change would not impact design basis accidents or create new accidents since it is bounded by the analysis for flooding of the Circulating Water Screen House in USAR D3.6.3.6. New equipment malfunctions are not expected from this change since the applicable SX division will be inoperable and seismic

, qualification of piping, regarding freeze seal installation (weight) is bounded by the temporary shielding allowances. In addition, WS system failure due to flooding would be enveloped by the Loss of AC Power event (USAR Section 15.2.6).

INSTALLATION OF TEMPORARY HOSES TO CROSS CONNECT SX AND WS Document Evaluated: TM 97-032 Log Number: 97-081 This Temporary Modification (TM) will cross connect the Plant Service Water (WS) and Shutdown Service Water (SX) Systems in a cc.nfiguration contrary to the USAR description.

This TM is needed to support the freeze seal for the interface valve between WS and SX, ISX014 A, evaluated in safety evaluetion Log Number 97-076. Up to four hoses will be installed to cross connect the WS and SX in order that the pressure between WS and SX can be equalized and the leakage past ISX014 A reduced. Once leakage is reduced, the previously evaluated freeze seal can be established. The WS system is designed to supply the Division 1 SX loads when

i l

Anachment 2 i t2 U 6028 M Page 106 of142 Divisien 1 SX is not available. This change does not impact the Ametionality of the WS System.

As such, the change would not change the impact of design basis equipment malfunctions. If the cross-connected hoses break, the manual isolation valves can be shut to isolate the leak or break.

In addition, the WS systen. has sufficient excels capacity to make up leakage in the event that the hoses fall. This change does not introduce any new accidents that have not be evaluated under the fteeze seal evaluation. No margins of safety as defined in the Technical Specification Bases are affbeted by this freeze seal. '

REMOVE REQUIREMENT FOR CONTINUOUS SODIUM MONITORING Document Evaluated: USAR Section 10.4.1.5,4, Table 9.3 3 Log Number: 97 082 This change removes the requirement for continuous sodium monitoring at the discharge of the  ;

comiensate pumps and the effluent of the tondensate polishers. These monitors only provide an  ;

indication of a tube leak. In line conductivity monitors in the condenser tube trays and the discharge from the condensate pumps provide the primary identifier of tube leaks. Sodium ,

monitors were primarily to be used for leak detection at plants that used aea water or bracLinh f water for condensate cooling. Their practicality for use in fresh water plants is diminished due to the lower qusntitles of sodium present. In addition, conductivity remains the most sensitive and reliable method of condenser leak detection at CPS and the only one described in USAR Section 10.4.1.4.

W process sampling system is classified as a non safety r' e lated, non-seismic category I syster*

in accordance with USAR Section 3.2. The in line conduuivity monitors for the tube trays  !

provide the first line of defense for the detection of a tube leak. The tube trays provide a sample that is not diluted and thus provides a more concentrated sample for leakage detection. hoe monitors are nere reliable since a more concentrated sample is used. As such, removal of sodium monitors would not impact design basis equipneent malfunctions. Accidents involving the i presence or absence of the sodium monitors is not described in the USAR: therefore, USAR accidents are not affected by this change Since conductivity monitors falfill the function ofleak detection that sodium monitms provide, the change would not create new accidents or equipment malfunctions. The removal of the monitors would not be a decrease in the margin of safety since more reliable equipment is being used.

i BREATHING AIR SYSTEM TAGOUT Document Evaluated: Tagout 94-1085 Log Number: 97-083 ,

Tagout 94 1085 removes from service the Breathing Air (RA) compressors ORA 01CA and ORA 01CB due to the equipment not being used for reliability reasons. The compressors slong with the support components such as piping, isolation valves, filter, heaters and associated annunciators, described in USAR Section 9.3.1, are also not being utilized. Portions of the system which penetrate the containment and drywell walls are Seismic Category 1, Class 2 design.

The portion of the RA system pressurized by the RA compressors provides air independent of the Service Air (SA) and Instmment Air (IA) Systems to various areas of the plant for maintenance 4

activities. The Main Control Room (MCR) emergency air bottles supplied by the fill station are d

.,a-r-- , .- = ,- - . - - . - - - - ..-.-.~,-,-n-- --.--n-c , n-...<c~ ,---~---,,----,.n.-.- , - - , , , . . , . - - , - - + ~ ~ , . - - -

[ Anacluneet 2 O U 602836  :

Page 107 of142 [

i not asected by this change. In addition, a portable system has be obtained which purines SA - l j supply air utilizing it to replace the RA system breathing apparatus ibnctions.  ;

The safety related portions of the system (containmnt and drywell isolation) will remain operable  ;

! and be periodically tested in accordance with Technical Specification surveillance requirements.

However, the normal position of the RA valves (with the exception of MCR related portions) will }

be changed from normally open to closed. All safety niated portions of the RA system, such as t i

the seismic qualified areas, are unarected by this change; therefore, there would be no impact to -

design basis equipment malfbnctions. Since the safety functions of this system are unaffected by i

this change (i.e., containment isolation, MCR breathing air bottles and Seismio Category 1 piping), this change would not impact design basis accidents. As safety functions will still be t maintained, the change would not introduce new dealgn basis accidents or equipment I

malfunctions. The safety fimetions of the RA system wiS. remain intact. The remainder performs .

no functions to maintain protective barriers. Therefore, this change would not decrease the margin of anfety.

INSTALLATION OF TEMPORARY HOSES TO CROSS CONNECT SX AND WS ,

Document Evaluated: TM 97-033 . LogNumber: 97-084 This Temporary Modification (TM) will cross connect the Plant Service Water (WS) and Shutdown Service Water (SX) Systems in a configura'.lon contrary to the USAR description.

- This TM is needed to support the fteeze seal for the interface valve between WS and SX, 4

.1SX014B, evaluated in safety evaluation Log Number 97-076. Up to faur hoses will be installed

to cross connect the WS and SX in order that the preasure between WS and SX can be equalized

[ and the leakage past 1SX014B reduced. Once leakage is reduced, the'previously evalua4d freeze seal can be established. The WS syatem is designed to supply the Division 2 SX loads when i 4

Division 2 SX is not available. This change does not impact the functionality of the WS System. l j As such, the change would not change the impact of design basis equipment malfunctions. If the cross-connected hoses break, the manual isolation valves can be shut to isolate the leak or break.

In addition, the WS system has sufficient excess capacity to make up leakage in the event that the t

hoses failed. This change does not introduce any new accidents that have not be evaluated under i the freeze seal evaluation. No margins of safet) as defined in the Technical Specification Bases l are affected by this freeze seal.

j' IB21.F024'A VALVE STEM LEAKAGE ALARM POINT.5tYPASSED ,

i Document Evaluated: Tagout 95-0566 Log Number: 97-085 i This change bypasses the high temperature alarm point on leak detection system recorder IE31 F616. This alarm is associated with 1B21 F022A (main steam line "A" inboard isolation valve).

W valve stem leakage system, as discussed in USAR Section 7.7.1.24.10.1.7, is composed of a 1

leak offline, alarm recorder and the ability to isolate the leakoffline. The recorder alarm point j has been bypassed ~due to a malfunction in the alarm circuit. It has been decided not to repair this; l instead, ECN 9511 will be implemented to install Chesterton packing and abandon the leakoff line. The work will be performed during the next window of opportunity. For ALARA reasons, 1

_- _ _ . _ . _ _ . _. . . ~ ___ - _ _ _ . - - . _ .

Anachment 2 l G U 602836 Page 108 of142 this window of opportunity has been defined as when IB21 F0'12A needs further maintenance (packing failure or failed LLRT).

This change will not change design basis equipment malfunctions as the leakofflines are a maintenance aid to locate small valve packing leaks. The leakoffline is routed to the drywell equipment drain system which is the primary means ofleak detection. Without this alarm bypassed other incoming alarms would have the potential to be masked, in addition, this alarm is not the prinary leakage detection .nethod. As such, this change would not affect design basis accidents. Since this tagout does not affect the operaticn of equipment other than the alarm, the change would not create new accidents or equipment malfunctions. Operational leakage is discussed in the Bases. Ilowever, this change does not affect these hakage limits or the margin of safety.

DRYWELL SUMP PUMP CMACITY GREATER THAN 120 GPM  ;

Document Evaluated: USAR Section 5.2.5,4.1 LogNumber: 97-086 i

USAR Section 5.2.5.4.1 spec;fies a value of 120 gpm for the Daywell Equipment Drain (RE) and Drywell Floor Drain (RF) pumps. Actual pump capacities are 150 gpm. This safety evaluation evaluates corrective action associated with condition report 196 08101 that changes the USAR 5.2.5.4.1 firom 120 gpm to 100% capacity. This has existed since initial licensing of CPS.

Changing this wording would not affect any equipment important to safety. The leak detection function, as defined in USAR Section 7.7.1.24.10.1.1, is not ' changed by this. In a:idition, this change will make the wording consistent with USAR Section 9.3.3.2.2.a, " Sumps and Sump Pureps," which sistes, "For general drainage each sump is fitted with two 100% capacity sump pumps." As such, this would not affect design basis equipment malfunctions. The RE and RF systems are not safety-related. There is no sailure analysis associated with these systems, Therefore, this change would not impact design basis accidents. Since no physical changes are ,

being made to the plant, no possibility for new malfunctions or accidents would be created. Since these systems are not safety related, there would be no impact to margin of safety by the wording change.

REPLACE COMPONENT COOLING (CC) SYSTEM RELIEF VALVE (GL 96-06 ACTION)

Document Evaluated: ECN 30167 Log Number: 97-087 ECN 29911 resulted in the installation of a relief valve 1CC318 on the Component Cooling (CC)

System water header ICC42D-6 hi containment. The relief valve relief pressure was 140 psig.

The relief valve was installed in response to Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Ba ns Accident Conditions," concerns on overpressurization of containment penetrations due to thermal expansion following a post-design basi:; LOCA heatup. The installation of the relief valve was evaluated by Safety Evaluation Log Number %-074 and reported earlier in this report. ECN 30167 replaces the 140 psig relief valve with a 200 psig relief valve. The investigation for Condition Report (CR) 1-97-01-103 determined that the CC system is operated at 120 psig. The relief valve was replaced to minimize valve cycling due to normal pressure fluctuations and the closeness of the setpoint of the presious l

m.___._._______ _ _ _ _ . _ _ _ _ ._.. _._.._._

Attachment 2 to U402836

'ase 109 of 142 valve to the header pressure. N 200 psig valve is installed in the same location, will relieve to the same place, and is constructed of the same material as the previous valve. Therefore, the

• L analysis of SE %074 remains valid.

REPLACE COMPONENT COOLING (CC) SYSTEM RELIEF VALVE .

l Document Evaluatori: ECN 30168 Log Number: 97-088 ECN 29912 resulted in the installation of a relief valve ICC319 on the Component Cooling (CC)

. System water header ICC46B 8 in containment. W relief valve relief pressure was 140 psig.

1 The relief valve was installed in respc noe io Generic Letter (GL) %06, " Assurance of Equipmcnt

- Operability and Containment Integrity During Design-Basis Accident Conditions," concerns on

_overprescurization of containment penetrations due to thermal expanslon following a post-design basis LOCA heatup. W installation of the relleivalve was evaluated t y Safety Evaluation Log '

4 No., SE %-075 and reported earlier in this report. ECN 30168 replaces the 140 psig relief valve 3 - with a 200 psig relief valce. The investigation for Condition Repoit (CP.) 1 97-01-103 l' determined that the CC system is operated at 120 psig. The reliefvalve was replaced to minimize

! - valve cycling due to normal pressure fluctuations and the closeness of the setpoint of the previous

[ valve to the header pressure. The 200 psig valve is installed in the same location, will relieve to the same place, and is constructed of the same material as the previous valve. Therefore, the j analysis of SE 96-075 remains valid.

! REVISION TO PERIODIC LOG DESCRIPTION IN USAR

Document Evaluated

USAR Sections 7.7.1." 12,7.7.1.7.5.3 LogNumber: 97 089 i '

This change clarifies use of Daily and Monthly Periodic Logs, and separates these logs from l- Section 7.7.1.7.5.3, " Performance Calculation Programs," description in the USAR. The description of these logs is being modified to include the flexibility to change the con'.ents of the

. log as needed. In addition, the inclusion of fuel burnup data into the monthly log is being deleted.

The Performance Monitoring System (PMS) is classified as, "not related to skfety." h logs are

, utilized for monitoring actions only and are not used for mitigating the consequences of an l accident. Additionally, these logs do not provide any control function. Therefore, this change p , would not impact design basis accidents or equipment malfunctions. The PMS is optically isolated from safety related equipment and as such would not introduce new design basis

- accidents, malfunctions of safety related equipment or decrease the margin of safety.

! REVISE DRAWINGS TO SHOW ISX020B AS A NORMALLY OPEN VALVE Document Evaluated: ECN 30158, LogNumber: 97-090

) This change revises system P&lDs and Operations Schematic drawings to show valve ISX020B j as normally open rather than closed. This is the divisional drywell chiller isolation valve. This j - change'would allow both the ISX020A & B valves to remain open during operation rather tlum one valve open (shown as ISX020A) and one valve closed (shown as ISX020B). This change is to resolve condition report 1 %-05-205 which identified that CPS procedures did not match the r

-mt-.-4----< -.w.,,v.,,,w,r.-nee- w ,---r,m+. ae ..eav. e- -e v.~~ +e*----e+v,-. v--w e -e m , - *-- -

W At% x nent 3

$; to U c4836 Page 110 of142 USAR. Revising the valve Icop to allow both valves open will provide plant operators, in using cooling water discharge valves, to automatically control cooling witer flow to the drywell coolers. The safety function of these valves is to close during an accident or low lake level to prevent a partial loss of dedicated shutdown service water to essential components. The drywell chillers are not safety related and water used to cool them does not return to the Ultimate Heat Sink (UHS) as other Shutdown Service Water (SX) System loads.

Since both vawes meet the same starJards, are tested and maintained equally, and are located on divisionally separate trains, the change would not increase probability of a malfunction of these valves. Equipment reliability problema are not expected with flows through the standby (shutdown) chiller unit. Because both these valves close on a LOCA signal, operation with both valves open would not affect design basis accidents. Loss of SX flow is already evaluated in the USAR. USAR Table 9.2-5 identifies SX system failures. Therefore, this change would not introduce new accidents or malfunctior <,. Since this does not impact the way this equipment is operated, the change would not impact margin of safety.

INSTALL A TEMPORARY PLUG IN THE DRYWELL FLOOR DRAIN INLET TO THE s WEIR BOX Document Evaluated: TM 97 034 Log Number: 97-091 This Temporary Modifht%TM) allows the installation of a i*echanical plug on the inlet piping to the drywell floor daMFm weir box during biocide treatment of the upstream piping. The plug will be installed dmEiF-b while in MODE 4 and 5 only and will aid in the cleaning of the RF piping in the drywell per maintenaace work request D70490. Provisions have been made to cllow a temporary weir box cover to be installed to prevent any damage to the capacitance probe during the draining evolution. The cover may be used to preclude splashing of water on the under-vessel cuity floor. The disabling of the drywell RF system and associated leak detection I for the floor drain system is a change to the facility as described in USAR Sections 5.2.5.2.l(1),

7.7.1.24.10.1.l(1) and 9.3.f This plug will be removed and the system returned to normal after the biocide soak and flush It is estimated that the plug will be installed for not more than twenty-four hours.

The Drywell Equipment Drain (RE) and RF Systems perform no safety-related functions. All safety-related equipment is located more than six inches off the floor, therefore, no safety-related related equipment would be affected if the ficor drains back up while the plug is installed. As such, this TM would not impact design basis equipment malfunctions. Since this system is not required in MODES 4 and 5, and the temporary plug must be removed prior to entering MODES 1 or 2, this would not create the potential for new equipment malfunctions or accidents.

Technical Specification 3.4.7 requires leak detection to be OPERABLE only in MODES 1, 2 and

- 3. Therefore, margin of safety would not be reduced by this TM.

E

~

Machn.aet 2 to U602836 -

Page 111 of142 CORRECTION TO PHYSICAL LOCATION OF AREA RADIATION MONITORS (ARMS)

Document Evaluated: USAR Sections 6.2.4.2,'et, al. LogNumber: 97-092 This USAR change revises locations of various' Area Radiation Monitors (ARMS) as described in the USAR The following ARMS have been changed to clearly identify actual areas monitored:

1RE-AR010,1RE-AR011 and IRE-AR013. The location of the following ARMS was changed to reflect actual locations within the plant; 1RE-AR013, IRE-ARO15,1RE-AR016 and 1RE- -

AR039. This change corrected locations of ARMS on the following drawings; 1RE-AR039 on Figure 12.3 5,= IRE-AR035 on Figure 12.3-20, IRE-AR013 on Figures 12.3-4 and 12.3-24, IRE-

' AR010 and 1RE-AR006 on Figures 12.3-10 and 12.3-25, and IRE-AR001, IRE-AR015 and

'1RE-AR016 on Figures 12.3-14 and 12.3-26.- Reference to ARMS within Residual Heat Removal (RHR) rooms "A" and "B" is removed from USAR Section 6.2.4.2 since no monitors are located within these rooms; Operational controls provide effective control of personnel exposure within these rooms; therefore, monitors are not required in thme rooms. For the remaining ARMS,

, ANSI /ANS-HPSSC 6.8.1-1981, " Location and Design Criteria for Area Radiation Monitoring hyncms for Light Water Nuclear Reactors," is followed and Regulatory Guides 1.97, .

i w umentation for Light; Water-Cooled Nuclear Plants to Assess Plant Conditions During and

% wing an Accident," and Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Exposures at Nuclear Power Stations Will Be As Low As Reasonably Achievable,"

are met.

The changes to areas monitored and monitor location are minor in nature. The areas are still monitored.- Only the monitor's location as depicted on USAR Figures is affected or the area monitored better defined. As such, these changes do not constitute sufficient change that in and -

ofitself would change, create or impact design basis equipment malfunctions, accidents or the margin of ssfety.

ARMS utilized prevent unnecessary or inadveitent personnel exposure. Sufficient controls exist to ensure that personnel safety is maintained. Both RHR "A" and "B" RHR pump rooms are

. posted as radiation and high radiation areas. In addition, the equipment in both rooms is located l* hind a shield wall to protect personnel from inadvertent exposure. Since there is no change to

- .quipment operation, malfunctions would not be affected. The ARMS are non-safety related nor are they required by the Technical Specification (TS), TS Bases or Operational Requirements Manual (ORM). As such, not providing these ARMS in these rooms would not impact the TS Bases margin of safety.

CORRECT THE REPRESENTATION OF INSTRUMENT AIR COMPONENTS IN USAR Document Evaluated: USAR Figure 9.3-2, Sheet 21 Log Number: 97-093

- USAR Figure 9.3-2, Sheet 21 is being revised to show the correct representation of the manual Instrument Air (IA) isolation valve IIA 229, air regulator IIA 17MG and solenoid air isolation IFSVQ138 relative to the drywell purge inlet valve, IVQO20. This change is being made to match the design configuration. As such, the change would not impact design basis equipment malfunctions or accidents. This change does not impact the operation of the IA system, Drywell

. Attachment 2 to U402836 Page 112 of 142 Purge (VQ) system, or valve IVQO20, As such, this change would not create new accidents or equipment malfunctions. This char.ge is being made to the non-safety related portion of the IA

- tubing and associated instruments, and therefore, would not reduce any margins of safety.

- CORRECT THE REPRESENTATION OF INSTRUMENT AIR COMPONENTS IN USAR Document Evaluated: USAR Figure 9 3-?., Sheet 6 Log Number: 97-094 USAR Figure 9.3-2, Sheet 6 is being revised to show the correct Instmment Air (IA) temperature instrumentation downstream of valve IIA 184. Temperature controller OTIC-WO756 will be -

- shown and not OTT-VWO74. Instmment OTT-VWO74 is already shown on this sheet in the correct locations. The representation of these controllers as contained in other design documents

- is correct. These instruments and their associated air handling units perform no safety related

- function' nor impact any safety-related function. As such, these changes will not impact design basis equipment malfunctions or accidents since physical plant configuration is unchanged. This change only impacts how the instruments are represented on the referenced drawings and would therefore not create new accidents or malfunctions. Since ; .ese instruments do not have any safety related function, there is no impact to any margin of safety.

CHANGE TO CGCS COMPRESSOR ROOM TEMPERATURE Document Evaluated: ECN 30077 Log Number: 97-096

. ECN 30077 revises the minimum operating temperature in USAR Table 3,11-5 from 35 to 40 'F based on calculation VR-57 R/1. This is the minimum temperature that the Combustible Gas Control (CGCS) System compressor rooms are required to maintain equipment qualification. The revised enviro.unental conditions are a result of not having cooling flow for the CGCS compressor rooms. This ECN also incorporated provisions to account for thermal expansion of fluid under LOCA conditions during design considerations. This was done to prevent recurrence of this potential condition (thermal overpressurization).

ECN 3007 also made editorial changes to the USAR. The High Pressure Core Spray (HPCS)

, System was identified as being in the Fuel Building in other USAR sections except for 3.4.1.1.

This section was revised to reflect HPCS in the Fuel Building. USAR Section 3.11.9.8 is also being revised to clarify that bounding environmental conditions are used to establish environmental qualifications (EQ) of equipment in the HPCS pump cubicle.

USAR Section 6.2.5.1.1 states that the CGCS is designed to meet Seismic Category I requirements. Also, the mixing compressors are designed to remain operable in the post accident environment in the containment building. The revised temperature on equipment in the CGCS compressor rooms was evaluated and equipment qualification is fully maintained. Therefore, this change would not impact or create new design basis equipment malfunctions. The probability and consequences of an accident (LOCA) have not been increased because potentially affected equipment has been fully qualified to the new parameters. For the same reason above, this change does not affect or create new design basis accidents. Since this change will not impact the

h 9 .6,== 2 to U 602836 Peps 113 of142 qualification of this equipment, protective barriert would not be affected, and the margin of safety has not been reduced.

I ADDITION OF VALVE lE51 F095 TO RCIC DESCRIPTION Document Evaluated: CR 1-97-05-252 Log Number: 97-097

. Condition Report (CR) 1-97-05-252 id tified a deficiency in USAR Sections 5.4.5 and 7.4.1.

These sections describe the operation of the Reactor Core Isolation Cooling (RCIC) System. The -

description for these sections does not include the Steam Admission Bypass valve (IE51-F095).

The steam Admission Bypass valve will, by design, open spproximately 10 seconds before the Steam Supply valve (IE51-F045). The bypass valve opeds before the steam supply valve in order to start and run the RCIC turbine to idle speed before the steam supply valve op;ms to bring it to full speed. During Test Loop Operation, the RCIC system performs a full flow test of the system by pumping water from the Condensate Storage Tank (CST) through the system piping and back '

to the CST. In Test Loop Operation, the Steam Admission Bypass does not open automatically L and manual opening is required. This change adds a description of the operation of the Steam Admission Bypass Valve to the USAR. The change to the USAR will increase awareness of the valve and ensure that it opens approximately 10 seconds before the Steam Supply Valve.

Opening of the steam admission valve before the steam supply valve will pre-warm the RCIC turbine and reduce thermal stresses on it. The steam supply valve does not impact other systems and as such the probability or consequences of failures of equipment important to safety will not increase. Addition of the description for operation of the steam supply valve will not introduce any failure modes or change the current operation of RCIC. Therefore, the probability or the possibility of accidents, previously analyzed or new types, is not increased. The Technical Specification Bans do not include the opening of the steam admission bypass valve, as such, the margin of safety is not decreased.

CORRECTION TO US AR FIRE PROTECTION (FP) DESCRIPTION P

Document Evaluated: CR 1-96-11-331 - Log Number: 97-100

- USAR Section 1.2.2.6.16 and Table 9.2-1 list the normal source of water to the FP system as the Service Water (WS) system. USAR Appendix E Section 3.1.2.6, USAR Section 9.5.1.2.2.2, and USAR Section 9.5.1.2.2.2 list the ultimate heat sin'< u the normal source of water to the FP system.: This USAR change achieves consistency between USAR sections by changing USAR Section 1.2.2.6.16 and Table 9.2-1 to reflect Lake Clinton as the normal source of water to the

'FP system. This change is an administrative, editorial change only. This change does not introduce hardware changes or operating parameters or new methods of operation. Therefore, this change will not have any impact on probabilities of failures of equipment important to safety, will not impact probabilities of accidents evaluated in the USAR, or create any new type of accidents not previously considered. This administrative and editorial change does not affed the Technical Specification Bases and as such the margin of safety is not reduced.

Attachment 2 to U 602836 i Page 114 of142 CORRECTION TO FLOOR DRAIN DRAWINGS Document Evaluated: ECN 30149 - Log Number: 97-101

~ ECN 30149 corrects discrepancies with drawings identified in Condition Report (CR) 1 97 107. The ECN corrects drain valve designators for three floor drain valves into which the Shutdown Service Water (SX) System thermal relief valves drain. USAR Figure 9.2-2 is affected by these drawing changes. In USAR Figure 3.6-1, the ECN also corrects the elevation designation for two drain valves in the Fuel Building and High Pressure Core Spray (HPCS) pump room Floor Drain (RF) System. The drawing changes are editorial in nature, no physical changes to the systems are made. The RF system is not safety related (USAR Section 9.3.3) and not required to shutdown the plant or to maintain it in a safe r.hutdown condition. The drawing changes affect the designations on the drawing, the drawing changes do not affect any safety related equipment and as such, the probability or consequence of failure of equipment important

, to safety is not increased As the RF system is non-safety and not required for safe shutdown or accident mitigation, the probability, or possibility of accidents new or previously evaluated is not

- affected. Drywell floor drains measure unidentified leakage and as such are mentioned in Technical Specification 3.4.7. This drawing change does not affect the Drywell floor drains and as such, this drawing change will not reduce the margin of safety.

RESERVE AUXILIARY TRANSFORMER PRIMARY TAP CHANGE Document Evaluated: ECN 30102 Log Number: 97-102 This ECN 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 AC Distribution system voltage.

This ECN, in conjunction with modifications AP-027, AP-028, AP-029, and AP-032, and ECN 30169, address the degraded voltage condition that was recognized and reported in LER 94-005.

Safety evaluation Log Numbers97-064, 96-015,97-069, 97-133, and 97-126, respectively, correspond to the modifications and ECNs. Additionally, ECN 30145 provides additional instrumentation to monitor the AC Distribution System. The modification performed by ECN 30145 was completed in July 1997, the t.ssociated Safety Evaluation will be reponed in a later revision to this report. These mcdifications and ECNs will ensure that adequate system voltage are supplied to all Class IE equipment. The ECNs and modificati_ons increase the RAT secondary

. voltage and change the facility as described in the USAR by; 1) Increasing the transient / steady state vohage analysis of USAR Section 8.2.2.1; 2) Increasing the fault currents outlined in USAR -

Tables 8.3-19 and 20; 3) Requiring balsucing of the 4.16KV loads and thus, change the description of plant operation as described in USAR. Sections 8.1.3.1 and 8.3.1.1; 4) Changing the

. plant loads that tap off the Main Power Transformers (MPT) as described in USAR Sections 8.3.1.1, and 8.1.3.1; because of the re-balancing of the 4.16KV loads; and 5) Designating the

. Emergency Reserve Auxiliary Transformer as the preferred power supply to the Clast 6 busses when shutdown and under a lightly loaded condition thus changing the description 6 plant operation of USAR Section 8.3.1.1.2. However, the final modifications to the degraded voltage issue will be implemented at a future date and will be reported in USAR Revision 8.

Anachment 2 to U-602836 Page 115 of142

The charge 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 1E equipment. Therefore, since all Class IE components are operated within their design ratings, the probability or consequences of failure of equipment important to

-~

safety would not increase. Generator load rejection is evaluated in USAR Section 15.2.2, a loss of any Main Power Transformer (MPT) is an initiating event for this accident. ECN 30145 installed MPT monitoring instrumentation to ensure that an overload condition does not occur at the MPT. Loss of AC Power accident is evaluated in USAR Section 15.2.6, the loss of an Unit

- Auxiliary Power (UAT) transformer, either by equipment failure or actuation of the transformer protective circuitry and 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. Therefore, the repositioning of the tap on the RAT would not increase the probability or consequences of any.

load rejection or loss of AC Power accident. This modification provides protection against

degradd voltage conditions while ensuring that IE loads do not exceed their design parameters.

This modification does not introduce any new failure mechanisms to the IE equipment it supplies.  ;)

Therefore, this modification does not increase the possibility or consequences of occurrence of any new type of accident not previously evaluated. The Technical Specification Bases discuss frequency and voltage specifications for the AC distribution system. As this modification does not violate these frequency and voltage specifications, the margin of safety is not reduced.

CHANGE TO USAR TABLE 12.5-2, PORTABLE AND LABORATORY TECHNICAL EQUIPMENT ANDINSTRUMENTATION Document Evaluated: USAR Change Package 7 Log Number: 97-103 USAR Table 12.5-2 is revised to cla.ify descriptions of portable or survey instruments. The

- changes correct types ofinstruments, location of storage, range ofinstruments, and quantity of _

pcitable or survey instruments available at CPS. The changes also correct minor typographical  !

errors in the table. No actual equipment is changed by this USAR change. The changes do not conflict with the requirements for the selectior, of radiological equipment contained in USAR Section 12.5.2. Similarly, the changes do not contradict the compliance wid. the Regulatory Guides and ANSI standards contained in USAR Section 1.8. These changes are primarily descriptive or editorial in nature and do not affect the probability of failure of equipment related to safety. Also, because of the natwe of the changes, they will not act as initiator or introduce any new failure modes that will impact any new or previously evaluated accident. Similarly, these cnanges, because of their nature do ne affect the basis for any technical specification and as such, the margin of safety is not reduced.-

USAR OPERATIONAL DESCRIPTIONS OF MOTOR OPERATliD VALVES (MOV)

Document Evaluated: CR l-97-04-062 Log Number: 97-104 Generic Letter (GL) 89-10 " Safety Related Motor Operated Valve Testing and Surveillance" required licensees to verify that MOVs that perform active safety functions are able to perform l

l

- Auachment 2 to U.602836 Page 116 of 142 --

[

_those functions when subjected to the maximum design conditions. CPS's analysis of MOVs under the GL 89-10 program yielded operational descriptions and design basis values that conflict with the USAR sections that describe those MOVs. This Safety Evaluation (SE) evaluates the inaccurate identification of active safety function MOVs in the US AR. _ The SE also evaluates information in the USAR that does not accurately reflect the worst case differential pressures that MOVs may be required to_ overcome to perform their active safety functions. Both the docketed response to GL 89-10 and the USAR are licensing basis documents (LBD). The new calculations under the GL provide different design basis values. The purpose of GL 89-10 was to provide assurance that active safety function valves perform their active safety functions. GL 89-10 does

- not affect the probability of failure of equipment important to safety. Since the GL 89-10 calculations did not change any plant hardware or introduce any new operating modes the probability of occurrence of accidents previously evaluated is 'not increased. Since there is no physical change to the MOVs and no new failure modes introduced. the possibility of a new type

of accident is not increased. The GL 89-10 program provided assurance that MOVs important to safety would perform their intended r,afety function. The GL 89-10 program, nor its results, arrected the basis for any technical specification. Therefore, the margin of safety is not decreased.

- USAR CHANGES ASSOCIATED WITH CR 1 %08-063 Document Evaluated: CR l-%08-063 Log Number: 97-106 This USAR change makes five changes to the USAR as corrective action for Condition Report (CR) 1-%08-063. The clanges are administrative in nature and add clarifications to the USAR.

The first change is to Table 5.2-6 and clarifies that Chemistry limits are specified in individual chemistry procedures. The second change adds a clarifying sentence to Section 10.4.1.1.2.3.

This clarifying sentence explains that oxygen addition to condensate and feedwater systems is to -

maintain oxygen levels as specified in that section. The third change is to Section 10.4.1.1.2.7; this change eliminates the reference to the regeneration of resin beds. CPS does not regenerate resin,- it replaces it when expended. The fourth change clarifies that dissolved oxygen is not monitored in the feedwater or condensate system when those systems are shutdown. The fifth change makes typographical corrections to USAR Appendix D.

The changes do not impact equipment operation nor add hardware to the plant. As such, these minor changes will not increase the probability of failure of equipment important to safety. As

- these changes are only administrative, the probability or possibility of accidents is not increased Similarly, and for the same reasons, the margin of safety is not decreased.

USAR DRAWING AND WO SYSTEM DRAWING DIFFERENCES Document evaluated: CR l-%11-041 Log Number: 97-107-Condition Report (CR) 1-%-11-041 identified differences between station drawings and the corresponding USAR figures. Drawing M05-1117-16 and USAR Figure 9.2-15 both had the incorrect line designator for line IWOB6B-3. Additionally, the connection between drain line IWO108A with valve IWO313 and valve IWOl93 was incorrectly shown on M05-1117-18 and on USAR Figure 9.2-15. The Plant Chilled Water System (WO) is described in USAR Section

Attachment 2 to U-602836 Page 117 of 142 9.2.8.3 and Table 9.2-21. The changes to the MOS drawings and the USAR Figures do not conflict with the system descriptions contained in the text or table. The WO System is non-safety related except for those portions located between and including the containment isolation valves

'and drywell isolation valves. The changes made to the USAR figures do not have any impact on -

the safety function of the system and as such the drawing changes do not increase the probability of failure of safety related equipment. The drawing changes do not prevent those components required for containment or drywell isolation from performing their safety function. The drawing changes do not act as initiators for any of the accidents previously evaluated and as such the probability or consequences of accidents previously evaluated are not increased. The drawing changes do not introduce any new failure modes or methods of operation. Therefore, the drawing changes do not increase the possibility of a new type of accident. The WO system valves that serve as containment or drywell isolation valves are governed by Technical Specification 3.6.1.3 and 3.6.5.3. The drawing changes do not affect the valves and as such, the margin of safety is not decreased.

INCORRECT LINE DESIGNATION ON BREATHING AIR SYSTEM LINE Document Evaluated: CR l-%-011-105 Log Number: 97-108 The Breathing Air (RA) System line designator for line ORA 02A-1" is incorrectly shown, as  !

ORA 32A, on USAR Figure 9.3-3 and drawings M05-1065 Sheets 5 and 6. The breathing air system is described in USAR Sections 9.3.1, Chapters 13 and 14, and Tables 3.2-1, 3.8-5, 3.9-5, 6.2-47 and 9.3-1, Changing the line designator in Figure 9.3-3 does not contradict the information in the text or tabies listed above. The breathing air system is non-safety related

except for those components.that perform containment isolation function. The change to the line designator is minor and will not affect the isolation functions of the system. The line, ORA 02A, routes RA piping from the auxiliary building to the fuel building, and the change to the line designator will not affect the containment isolation functions required of the system; therefore, the change in line designator will not incraase the probability or consequences of failure of equipment important to safety. The change in line designator does not affect the accident analysis and as such will not increase the probability or possibility of any accidents, new or previously evaluated. This change to a line designator does not affect any bases and as such the margin of safety is not decreased CHANGE TO USAR SECTION 7.6.1.3.3.1 Document Evaluated: CR 1-96-09-141 Log Number: 97-109 USAR Section 7.6.1.3.3.1 used to state that check valve position can be confirmed for Residual

.:st Removal (RH) shutdown cooling return line check valves IE12F050A/B at any time. - These valves IE12F050A/B are tilting disc check valves that have no limit switch or mechanical position indicators. The USAR has been revised to remove this statement The purpose of the RH shutdown cooling return line check valves (IE12F050A/B) is to protect the RH piping from the higher pressure reactor vessel pressure. The RH system is widely discussed in the USAR and Technical Specifications. Removal of the statement that the valve

i-l l;

Anachment 2 to U.602836 Page 118 of142 rosition can be confirmed does not contradict any other section of the USAR or Technical Specifications. Removal of the statement will not affect system alignment or operability of the RH

system. Removal of the statement will not impair the valves or the RH system firom performing their design or safuy functions. Revision of USAR Section 7.6.1.3.3.1 to remove the statement will not have any effect on the valves or the system. Therefore, removal of the statement will not

-increase the probability of failure of any equipment important to safety. All safety functions of the RH system continue to be met, therefore, the probability of an accident previously evaluated are not increased. Removal of the statement will not introduce any new failure modes or operaths modes. Credit is not taken in the safety analysis for the check valve position, and therefore, removal of the statement will not increase the possibility of a new and unanalyzed accident. All technical specification requirements of the RH system will be met after removal of the statement.

- Therefore, the margin of safety is not reduced.

ACOUSTICAL CEILING TILE FLAME SPREAD AND SMOKE DEVELOPMENT RATINGS Document Evaluated: CR 196-07-074 Log Number: 97-110 Rl' The acoustical ceiling tile flame spread and smoke development ratings of USAR page E4.1-18 (Item f) and USAR Appendix E Section 3.l.2.2.7 are not consistent. Both USAR sections  ;

conflict with vendor-supplied information. This change evaluates revision of the USAR to incorporate the vendor-supplied information. The vendor data supplied with the ceiling tile is equal or superior to that contained in both USAR section references _h ceiling tile does not affect any equipment important to safety other than by fire. In a fire, the tile would act as fuel not as ignition source. The fire would spread slower and the smoke contribution would be smaller with actual vendor supplied as-built information Therefore, revision of the USAR sections will not increase the probability of failure of equipment important to safety. Similarly, as the tile

- would act as fuel and not as ignition source the probability of a fire is not increased With the fire ratings superior to those in the USAR, the fire would spread slower and the consequences of the fire would not increase. The tile cannot initiate any credible type of unanalyzed accident, and thus the possibility or consequences of an accident not previously evaluated are not increased The Technical Specification Bases do not contain reference to fire protection and thus the incorporation of vendor supplied fire ratings into the USAR would not decrease the margin of safety.

DRY SOLID WASTE PACKAGING EQUIPMENT Document Evaluated: USAR Section 11.4.2.5 Log Number: 97-113

.. USAR Section 11.4.2.5 describes the general aspects of dry active waste operations. The use of-

'offsite processing has become the standard mode ofprocessing dry active waste. The use of offsite dry active waste processink involves temporary storage of the dry active waste. This change to the USAR addresses the temporary storage of dry active waste outdoors but within the protected area. The temporary outdoor storage location for dry active waste is on the asphalt and gravel area outside the turbine building roll-up door. The storage location is at a higher grade than the surrounding area to ensure no water accumulation. Safe operating distances are

Anachment 2 to U 602836

(_ Page 119 of142 maintained to ensure containers do not come in contact with the pylons of the Reserve Auxilia y Transformer (RAT) during movement of dry active waste in or out of the storage area.

- The probability of failure of the RAT due to the movement of dry active waste into or out of the temporary storage area is minimal. Therefore, the pr'obability of failure of equipment important to safety is not increased. Three accident scenarios were considered in this safsty analysis. A fire, a seismic event, and a tornado were considered as accident scenarios. W analysis concluded that the conse<1uences of none of these accidents would be increased due to the storage of the dry active waate, h storage of dry active waste does not act as initiator for any of these accidents.

Therefore, the probability of accidents previously evaluated or the consequences of those accidents are not increased. Since the storage area is above the grade for the maximum flood of USAR Section 1.2.2.1.2.2, a flood was not considered. The outdoor storage of dry active waste does not create any credible conditions that would initiate a new type of accident not previously evaluated. W radiological release limits of Technical Specification 5.5.4,10 CFR 20, or 10CFR50 Appendix I are not violated by the outdoor storage of dry active waste and the margin of safetyis not decreased UPDATE USAR FIGURE 6.5-2 " SECONDARY CONTAINMENT LEAKAGE AS A FUNCTION OF WIND" Document Evaluated: CR l-96-10-185 Log Number: 97-114

~ Calculation VG-01 determines leakage rates from secondary containment at different wind speeds. The calculation verifies equipment design capacity and integrity of the secondary containment. In 1985, revision 1 of the calculation was added to the USAR Section 6.5,1,1,1, however the figure in the calculation was never revised in the USAR. This change revises the USAR Figure 6.5-2 with'the figure in Revision 2 of the calculation. Since the text of USAR

- Section 6.5.1.1.1 already corresponds to the text of the calculation, the revision of the US AR figure becomes administrative. This USAR change does not impact any safety related equipment and thus the probability of failure of equipment important to safety is not increased The revision of the figure does not impact any accident previously evaluated and as such, the probability of occurrence of design basis accidents is not increased Since revision of the figure does not affect operation of the ventilation system, the possibility of any new type of accident is not increased.

The USAR figure is revised to match the USAR text. There is no impact to the Technical Specification Bases by this revision to the USAR, and the margin of safety is not decreased CHANGE CONTROL ROOM VENTILATION VC PEAK HEATER REQUIREMENT Document Evaluated: CR l-97-04-226 Log Number: 97-115 h peak heater requirement of USAR Section 6.5.1.2.2.d.2 is revised to reflect the values ofIP Calculation VC-10. Heaters OVC02AA and AB are Control Room (VC) Make-up ventilation system air filter unit heaters. The function of heaters is to maintain the humidity of the air entering the control room to no more than 70 %. The lower humidity increases the efficiency of the make-up filter charcoal. The heater rating and capacity is not affected. W calculated peak heater requirement is less than that in the US AR, there is no adverse impact on the control room l

Anachaient 2 to U4028%

Page 120 of142 -

, ~ equipment, and the system performance is not adversely affected. Since the stated heater demand is lower and the capacity of the heater remains unchanged, no physical changes to the heater are required.- Therefore, the probability of failure of the heater or any equipment important to safety is not increased. Since the capacity of the heater is not changed, and there are no physical changes to the heater or the VC system, the probability or possibility of accidents or their consequences is not increased. Since no changes are made that impact the basis for any technical specifice. tion, the margin of safety is not decreased.

USAR CHANGE FOR THE INDEPENDENT ANALYSIS GROUP (IAG)

Document Evaluated: USAR Section 13.1.1.2 & Figure 13.1-1 A LogNumber: 97-116

- This USAR change adds the Independent Analysis Group (IAG) to the corporate structure relative to CPS Figure 13.1-1 A, "IP Corporate Structure Relative to Clinton Power Station," and USAR Section 13.1.1.2, " Organizational Arrangement." Additionally, USAR Section 13.1.1.2 has been revised to state that the Vice President reports to the Senior Vice President. The formation of the IAG is fundamentally an administrative change. The root cause investigation activities the IAG will be performing are currently performed by various groups or positions within the Nuclear Program under the generic Corrective Action Program. Those functions will continue to be paformed using IAG for assistance. The IAG will perform advanced root cause analysis using a focused team ofindividuals working under existing generic programs. Since the IAG's activities are controlled under existing programs which are not directly related to equipment manipulation, there is no impact on equipment malfunctions. The primary function of the IAG is to assist other site departments using enhanced root cause analysis techniques under the existing Corrective Action Program _. Therefore, this change would not affect design basis

' accidents.~ The administrative activities performed by the IAG would not affect any Technical Specification Bases or margin of safety. The change of who the Vice President reports to is a title change not affecting safety since the duties, responsibility and authority of the Vice President remain the same.

OPERATION OF THE RADWASTE BRIDGE CRANE FROM AUXILIARY CONTROL PANEL-Document Evaluated: USAR Section 11.4.2.7 LogNumber: 97-117 The USAR is being revised to describe the use of the auxiliary control station to operate the radwaste bridge crane when dose rates of the load and load sighting capability allow. Using the auxiliary control station for operation of the crane allows for safer load moves because the load is in direct observation of the crane operator. The previous USAR description ofihis control station stated that it is used primarily for crane maintenance and testing. This imyroved method allows the crane operator to directly oburve the load during rigging and movement. The only accident identified relative to this change would be a dropped liner with subsequent release. This accident requires a failure of the crane or rigging. The operation of the crane from the auxiliary control station may decrease the likelihood of rigging failure by the operator being able to directly observe the rigging during installation and load movement. The decision to use the auxiliary control station takes into consideration the dose rates from the package to evaluated ALARA

Attaciunent 2 to U402836 Page 121 of142 aspects of the lift. ; Since this change should minimize the potential for errors, the probability of an equipment malfunction important to safety has not been increased.

REVISED POSITION ON COMPLIANCE TO REGULATORY GUIDE 1.11 Document Evaluated: USAR Sections 1.8, et. al. Lc > Number: 97-118 This USAR change provides clarification ofIllinois Power (IP's) position regarding Regulatoiy Guide (RG) 1.11, " Instrument Lines Penetrating Primary Reactor Containment," and the use of Excess Flow Check Valves (EFCVs) at CPS. This change also resolves inconsistencies in the USAR on the type ofinstrument lines that contain EFCVs. EFCVs are flow restricting devices that are installed in instrument sensing lines that penetrate primary reactor containment. The low pressure EFCVs (20 total) are used in instrument sensing lines connected directly to containment

- _or drywell atmosphere. The high pressure EFCVs (2 total) are used in sensing lines connected directly to the reactor pressure boundary. All are located inside secondary containment less than ten feet from the containment penetration. Guidelines in RG 1.11 require EFCVs to actuate if there is a break in the downstream instrument piping both during nonnal reactor operation and following a design basis accident. USAR Section 1.8 adds an exception that the low-pressure EFCVs at CPS are not designed to actuate close on a failure of an instrument line during normal operation, when the pressure differential between primary containment and secondary containment is very low and the radiological source term is negligible. USAR Section 6.2.4.2 was revised to state that not only are EFCVs used in lines sensing suppression pool level, but they are also used in lines sensing drywell and containment ventilation pressures and the reactor coolant system. USAR Sections 6.2.4.3.2.4 and 7.1.2.6.3 were also revised.

In the event of an instrument line break outside containment, the excess flow created by this break is insufficient to actuate the ECFV, The functional limitation ofEFCVs to close during normal reactor operation has no safety consequence since there is negligible radiological source term available for release from containment or drywell. Any radiological release is further minimized by the relatively low flow through the instrument line. The function of an EFCV to close under accident conditions has not been affected, thus satisfying the intent of Regulatory Guide 1.11.

The change will not increase the consequences of malfunctions or accidents since, during normal reactor operation, the affects of an instrument line break are considered negligible. The inability oflow pressure EFCVs to close daring non-accident plant conditions will not impact any previously evaluated accidents. With such a low source term, although the check valve may not close, release to secondary containment would be insignificant and substantially below 10 CFR Part 100 limits. The EFCV is designed and tested to close at a differential pressure less than peak accident pressure inside containment or below the flow rate corresponding to the accident pressure. The acceptance limits placed on the EFCVs by the Technical Specification Bases and

- specifically called out in the Operational Requirements Manual are not reduced. Therefore, the margin of safety is not affected.

Attachment 2 to U 602836 L Page 122 of 142 CLARIFICATIONS TO CPS POSITION ON REGULATORY GUIDE 1.52

Document Evaluated: ANSI N509/510 Variance Report LogNumber: 97-119 This change clarified CPS' position with regard to the requirements of ANSI N509/510-1976 (ANSI N509-1976), " Nuclear Power Plant Air Cleaning Units and Components", ANSI N510-1975 " Testing ofNuclear Air Cleaning Systems." This standard is listed in Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filteration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," as a standard that can be used to satisfy the requirements of the general design criteria. The variance report, which is referenced in the USAR, lists exceptions to the requircments of ANSI N509/510 at CPS, This change defines and clarifies CPS's interpretation .

and soplication of ANSI N509, Section 5.5.1. This section indicates that the heater stage (to control humidity) shall be designed capable of reducing the entering air-steam mixture (100%

relative humidity) to approximately 70% relative humidity (RH). The entering air for the Standby Gas Treatment (SGTS) System at CPS is not an " air-steam," mixture. Air-steam mixtures would be present in atmospheric cleanup systems connected directly to the primary containment atmosphere. The entering air of the SGTS at CPS is considerably less than 100% RH. As such, this requirement is not directly applicable to the CPS design. This is supported by a later version of ANSI N509-1980 which states, "the heater stage shall be sized on the basis of heat transfer calculations showing its capability of reducing the maximum expected relative humidity of the entering air stream mixture to approximately 70%".

The design of the heaters under abnormal conditions will be based on the maximum expected relative humidity under worst case operating conditions (including degraded soltage). Since this

- equipment is designed to operate ur.:ler all conceivable worst case operating conditions, this change will not increase the probability of design basis accidents or equipment malfunctions. The SGTS is a mitigative system to limit the impact of an accident. Therefore, this change would not create new design basis malfunctions or accidents. Since the SGTS system heaters are designed to operate under worst case scenarios, the Technical Specification Bases and margin of safety would not be reduced.

ERAT TAP SETTING CHANGE FROM TAP #3 TO TAP #2 Document Evaluated: ECN 30169 Log Number: 97-126 This ECN repositions the primary winding tap for the Emergency Reserve Auxiliary Transformer (ERAT). This ECN,in conjunction with modifications AP-027, AP-028, AP-029 and AP-032 and ECNs 30102 address the degraded voltage condition that was recognized and reported in LER 94-005. Safety Evaluation (SE) Log Numbers97-064,96 015,97-069,97-133, and 97-126, respectively, correspond to the above listed modifications and ECN. These modifications and ECN will ensure that adequate system voltages are supplied to all Class IE equipment. This change is similar to the change made for the Reserve Auxiliary Transformer (RAT). The final modifications to the degraded voltage issue will be implemented at a future date and will be o reported in USAR Revision 8.

Auachment 2 -

to U-602:36 Page 123 of 142_--

The change in the ERAT secondary tap settings alone results in a decrease in voltage to Class 'lE equipment. Station calculations have further determined that when the ERAT transformer taps are changed as part of the implementation of the above listed modification and ECNs, IE components would not see an overvoltage condition. Therefore, since all Class IE components, are operated within their design ratings, the probability or consequences of failure of equipment important to safety would not increase. Loss of AC Power accident is evaluated in USAR Section 15.2.6, the loss of an 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 ERAT is designed to provide rsted power at all the tap settings, and would not cause the transformer protective circuitry to actuate. The UAT: and Main Power Transformers

. (MPT) are still operated within their design ratings. Therefore, the repositioning of the tap setting on the ERAT would not increase the probability or consequences of a loss of AC Power accident.

This modification provides protection against degraded voltage conditions while ensuring that all -

- IE loads are not operated outside of their design parameters. This mod;fication does not introduce any new failure mechanisms to the IE equipment it supplies. Therefore, this modification does not increase the possibility or consequences of occurrence of any new type of accident not previously evaluated. The Technical Specification Bases discuss frequency and voltage specifications for the AC distribution system. - As this modification does not violate these frequency and voltage specifications, the margin of safety is not reduced.

USAR DESCRIPTION OF REPROCESSING / DISPOSAL EQUIPMENT DRAIN (WE)

Document Evaluated: USAR Sections 1.2.2.9.2 & 11.2.2.1 LogNumbx 97-128 Condition Report 1-96-11-171 documented discrepancies between the USAR Section 1.2.2.9.2 and 11.2.2.1 descriptions of the_WE system operation and that provided by CPS procedure 3901.02, " Operating Equipment Drain Collection System." The USAR is revised to reflect CPS procedure 3901.02 to make sampling of the waste collection tanks optionst by stating that if sampling is required, tank contents are recirculated fo a minimum of 30 minutes prior to -

1 sampling. Once the sample is drawn, processing is recommenced. This procedure reflects the

- established consistency of collected batches and the accumulated operating experier.cc.

The waste water processing system is contained within plant buiidings such that any physical failure of components would not result in release of water to the environment.- Deleting sampling requirements would not affect or create design basis equipment malfunctions. Sampling to determine appropriate processing methods is not a factoi in accident evaluations. Therefore, substituting conservative selection processes, combined with post-process sampling would not

' impact or create new design basis accidents. Performance ofpre-processing does not factor into any margins of safety.

MOISTURE SEPARATOR REHEATER (MSR) BLANKETING STEAM Document Evaluated: Tagout 92-1572 Log Number: 97-129 USAR Sections 1.2.2.8.13 and 9.5.9.2 state that steam produced by the auxiliary steam boilers is used during start-up and shutdown to provide blanketing steam to the Moisture Separator

Attachment 2.

to U.602836 Page 124 of 142 Reheaters (MSRs). Additionally, USAR Table 9.2 1 lists the steam blanketing drain cooler.

Operations Tagout 92-1572 isolates all valves associated with MSR blanketing steam. CPS

. procedures are also being redsed (safety evaluation Log Number 96-058) to rerr.ove steam blanketing due to it being non-functional. The future use of steam blanketing system will be

- decided based on low pressure (LP) turbine inspection in RF-6 and the decision to replace LP rotors if needed. The replacement, if needed, is tentatively planned for RF-8. The steam blanketing system is designed to prevent rust from forming in the MSRs. The steam blanketing

_. system is presently non-functional. - This safety evaluation is to address removal of steam blanketing.

As defined in USAR Section 9.5.9.3, the auxiliary steam system has no nuclear safety-related components. Failure of this system would not impact safe operation of the plant or increase the -

probability or consequences of equipment malfunctions or accidents. Also, this change would not create the possibility of new type of equipment malfunction or accident. As stated in 9.5.9.3, this system does not have safety-related components and the change does not reduce any margin of safety.

. SERVICE BUILDING HUMIDIFICATION BOILER (0VS19B) REMOVED FROM SERVICE Document Evaluated: Tagout 96-0233 Log Number: 97-130 Operations Tagout %-0223 removes the Service Building HVAC (VS) humidification boiler

_ (0VS19B) from service. This boiler supplies Service Building humidifiers (OVS06M, 0VS07M, OVSO9M and OVS24M) shown in USAR Figure 9.4-20,- Sheets 1 and 2. Boiler OVS19B was -

- removed from service for maintenance purposes. This unit provides humidity control to the service buildins during the winter months. There are no plans to restore this equipment to service in the near future. - All other VS equipment iemains unaffected by this tagout. This svstem is only intended for personnel comfort, The VS system performs no aafety-related or design basis functions. The operation of VS in not required to: maintain the reactor coolant pressure boundary, shutdown the' plant or maintain the plant in shutdown conditions. Therefore, this tagout would not impact or create new accidents or equipment malfunctions. In addition, the VS system is located entirely within the Service Building which has no safety-related stmeture=, '

- Therefore, this change would not reduce any margin of safety for any technical specification.

PERMANENTLY CLOSE HALON PURGE DAMPER, OVS43Y Document Evaluated: ECN 30195 Log Number: 97-131 This ECN permanently closes the Service Building record storage facility halon purge damper OVS43Y by disconnecting the instrument air supply to the damper. This damper had been used to allow halon purge to facilitate reentry to the record storage facility. This change is necessary since an interlock exists between the damper and halon initiation which would allow halon to escape, jeopardizing fire suppression and violating National Fire Protection Association (NFPA)

_ Code 232. This purge is needed for personnel to return to the record storage facility following a halon initiation.- The purge function will be provided by use of temporary fans and flexible ducts.

This change would not impact equipment performance since the VS performance is unchanged.

i-Attachment 2 -

to U4028%

l

-* Page 125 of142 USAR Section 9.4.12.3.b states that, "A failure analysis is not presented since the service building HVAC system is not safety-related." Therefore, this change would not impact design basis

. accidents or create new accident or equipment malfunctions. _ Since this purge affects non-safety related equipment, there would not be an impact to any margin of safety.

4.16 KV BUS DUCT THROUGH FIRE BARRIERS CB-If7CB-Sa Document Evaluated: CR l-96-06-056 LogNumber: 97-132 Condition Report (CR) 1-96-06-056 identified a 4.16 KV bus duct through the fire barrier

between fire zones CB-If and CB-Sa in the Control Building. There had been no evaluation or deviation which accounted for the 4.16 KV bus duct feeds that pass through this fire barrier. The equipment associated with this change are the safety related equipment and cables located in fire

_ zones CB-If and CB-Sa that would be required to assure safe shutdown in the event of a fire, as discussed in USAR Appendix E, Section 3.4.1.6 (CB-1f) and 3.4.5.1 (CB-Sa). To preclude the possibility of a single fire affecting Division 1 and 2 cables, Thermo-Lag protected Division 2 cables, in conjunction with an hourly fire watch, protect this zone until permanent '1 hermo-Lag resolution is implemented. Zone CB-Sa cables were modified (Safety Evaluation (SE) Log No.

%-019) to remove reliance upon Thermo-Lag for Division 2 cables.

Since safe shutdown of the plant is assured by safe shutdown method 2 (using Division 2.

equipment), there is no increase in the probability of equipment malfunctions of equipment important to safety. There are twenty (20) 4.16-kV and 6.9-kV bus duct penetrations in 3-hour fire rated barriers. The bus duct side of the penetrations is constructed of 0.125-inch thick steel construction bolted to the barrier on each side of the penetration. As evaluated by fire protection engineering, this configuration provides equivalent protection as that provided by a fire door or damper, therefore, no new type of equipment malfunction would be created. Since safe shutdown can be achieved using method 2, Division 2 equipment for a fire in either area, barrier integrity would not bejeopardized.

REPLACEMENT OF REGULATING TRANSFORMSR WITH NON-REGULATING TRANSFORMER AND TRANSFORMER TAP Document Evaluated: Modification AP-032 Log Number: 97-133 Modification AP-028, Safety Evaluation (SE) Log Number 96-015, stalled new undervoltage

- relays and new regulating transformers to address degraded voltage concerns. AP-028 had additional changes that were also evaluated in SE 96-015. 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 Modification 97-017. This modification also makes changes to the tap settings in the distribution transformers inside the Motor Control

~ Centers OAP55EB, 0AP57E, and 1 AP75E. The final modifications to the degraded voltage issue will be implemented at a future date and will be reported in USAR Revision 8.

Attachment 2 l to U 602836 Page 126 of142 Additional changes were made by this modification, these changes, internal to the MCCs, switched individual loads from the non-regulated distribution panels to the regulated distribution panels. The changed loads retained 6.e divisional and physical separation of channels described in USAR Sections 7.6.1.2.5.1 ad i1.5.2.1.6. Illinois Power (IP) calculation 19-AI-62 verified i

proper breaker coordinckn and the supply breaker to the regulated distribution panel was verified to be adequate for the additional load. 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 Numbers97-057 and 96-015 remain valid for the re-connection of the non-regulating transformers 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 Transformer (RAT) and Emergency Reserve Auxiliary Transformer (ERAT). The changes to the RAT and ERAT were accomplished with ECNs 30102 and 30169 (SEs97-102 and 97-126 respecnnsy). Because of the above reasons, these changes do not increase the probability or conse mes of failure of equipment important to safety.

SE Log Number 97-057 evaluated the replacement of the regulating transformers with the previously abandoned _non-regulating transformers and determined there was no impact to the accidents previously evaluated. Station calculations 19-AJ. 72 and 19-AN-19 determine that the -

changes, internal to the MCC, assure the Division 2 loads, down to the 120 volt level, have sufficient voltage to operate. The design complies with GDC 17, therefore, the probability or consequences of occurrence of an accident previously evaluated is not increased.

Loads internal to the MCC, which required a regulated supply, were switched to the regulated distribution panel as the non-regulated transformers replaced the regulated transformers. Power supplies for various loads were changed to avoid the adverse effects of the higher voltages.

Damage to loads requiring regulated power supplies is avoided by this modification. This -

modification does not introduce any new failure modes as the internal changes to the MCC use

, the same equipment that existed previously. Therefore, the possibility or consequences of a new type of accident not previously evaluated are not increased.

Technical Specification (TS) Bases 3.3.8.1 were changed to support AP-028. This modification does not compromise those bases. TS Bases 3.8.9 and 3.8.10 require distribution systems to provide the capacity, reliability, and redundancy to support Engineered Safety Feature (ESF) systems, and other plant systems. This modification along with the others listed above restores the ability of the distribution system to meet those requirements. Therefore, this modification does not decrease the margin of safety.

RESOLUTION AND ASSOCIATED PROCEDURE CHANGES FOR CR l-96-07-074 Document Evaluated: CR l-%-07-074 Log Number: 97-134 USAR Appendix E, Item Id, " Applicant's Position," stated that the non-safety related drywell purge filter train deluge systems are connected to the Class IE power system. Contrary to this

__y . . - _ _ _ _. _ .. _ ~ . _ _ _ . _ _,. _ _.._.

Attachnent 2 to U-602636 Page 127 of142 i statement, the deluge valves for the breathing air filter (0FP261) and the drywell purge filter train l deluge system (0FP166A, B and C) were not connected to the station Clau IE power system.

The US AR has been revised to reflect that these components are not supplied by Class IE power i systems.

The USAR describes the ability of the fire protection system to respond to a fire, An engineering evaluation associated with this Condition Raport (CR) was performed to demonstrate that the deluge valves may'be operated electrically or manually in response to a fire. The manual deluge valves are located within five feet of the electrical switch. As such, anyone responding to a fire could manually operate the system as easily as using the switch. According to CPS procedure 5042.02, " Alarm Panel 5042 Annunciators - Row 2," an operator is dispatched, "to determine which DW Purge Train is overheating." Even though the control switch for operating each valve is located in the Main Ccontrol Room (MCR), the operator, who determines the exact location of the problem, could be directed to operate the appropriate valve manually since it is located outside the room containing the purge units. This evaluation determined that the deluge valves can fulfill their function even though they are not connected to Class IE power, therefore, the equipment failure analysis had the accident analysis are not impacted. Since the systems potentially affected by this are not safety-related, and bases for the technical specifications do not take credit for this equipment, this' change would not decrease margin of safety.

USAR TEXT AND DRAWING CHANGES IN RESPONSE TO EDSFI QUESTIONS Document Evaluated: .USAR Sections 8.3,1.4.2.4, et, al. LogNumber: 97-135 In response to NRC Electrical Distribution System Functional Inspection (EDSFI) queuion 621, the wording in USAR Section 8.3.1,4.2.4 is revised to remove listing of 100 Amp fuse protection of #12 AWG containment penetrations with DC circuit feeds. USAR Figure 8.3-9 is also revised for question 621 to delete the curve shown for Bussman fuse 25A, type KWN. In response to question 623, a 40-second time delay for loading the Main Control Room Ventilation (VC)

System fans onto the diesel generator after a loss of power event, was deleted. This time delay does not exist if the fans are already running. The 40-second delay is a part of the Auxiliary Power (AP) System design at the 480V Unit Sub level.

The correct 15A fuse continues to provide the required redundant protection for the type F containment penetrationt with DC circuits. The fuse curve was not being utilized for any function. In addition, the installed 40-second time delay in loading the fans onto the AC bus after a Loss of Offsite Power (LOOP) event continues to provide protection against excessive load

= from the VC system during the loading sequence. Thus, there is no change to design basis equipment malfunctions made by this change. This change clarifies various statements concerning the electrical design of the plant. This change does not impact design basis accidents as described in the USAR. No changes are made to the plant or the methods of operation. Needed protective equipment, functions and features are unchanged. Therefore, the change introduces no new accidents or equipment malfunctions. Since these changes clarify electrical design, and the type F containment penetration is still protected, this change would not impact the margin of safety.

Attadunent 2 ,

.to U 602836 '

1 Page 128 of 142 -

L CANCEL CPS PROCEDURE 6205.01 AND REVISE APPROPRIATE USAR SECTIONS -

Document EvaluatedF US AR Sections 8.3.1.4.2.4, et. al. LogNumber: 97-138 ,

, '.This cancels CPS procedure 6205.01," GAM RAD Fluid Analyzer (Turbidimeter)." This j- procedure is used for operating and testing the GAM RAD Model 370 A Fluid Analyzer. This

'i

! analyzer is installed in Feedwater (FW) Sample Panels IPL88JA/IPL88JB and feedwater

[ corrosion product monitor,1B21-Z001. This instrument is being removed from service but not g _ being physically removed from the applicable sample panels. Cancellation of this procedure -

L affects USAR Section 7.7.1.1.3.7 (3),7.7.1.1.5.2 (7),7.7.1.5.3, Table 9.3-3 and 10.4.6.5. This -

p analyzer v as designed to continuously monitor the condensate pump discharge header and FW for the presence of suspended matter (iron). The current method used for FW monitoring is the

, use of two types offilters installed in the FW corrosion product monitor. This method is in .

i accordance with the EPRI BWR Water Chemistry Guidelines.

USAR Section 7.7.1 states that the FW turbidity moni'or is not essential for the safety of the plant. In addition, monitoring is still being performed. As such, this change would not affect design basis equipment malfunctions. FW monitoring for suspended iron is not credited during a c design basis accident evaluated in the USAR. Therefore, this change would not affect design L - bas:s accidents. There are no credible new accidents or malfunctione associated with this non- .

I essential component change. The FW turbidity analyzers do not affect any acceptance limits;

therefore, the margin of safety is unaffected by this change.

i L TEMPORARY POWER TO TEMPERATURE CONTROLLER RACK IN PANEL IH13-P867 I

4 L Document Evaluated: TM 97-037 Log Number: 97-139

l -

[ - This Temporary Modification (TM) addresses one of the overvoltage concerns raised during the

! review of the circuits fed from the 120V distribution panel which are having their transformer taps

- changed to provide a 5% inorease. The power supply for the temperature controller rack in the j electronic control module portion of the H202 mcnitoring synem has limitations. on voltage j ir.put. 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 j of the air compressor for the H202 sampling panel (ICM0lSA) to energize the power supply for 1-the temperature controlle. rack in panel lH13-P867. This maintains the disisional power

, requirements of the equipment as well as the description of the containment monitoring system in j _ the USAR. However, schematic diagrams, listed in USAR Table 1.7-1, for the system are

imputed by the revised circuit wiring. This TM will be installed through operating Cycle-7 until permanent degrmied raltr.ge modifications are made, e

p 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 functiorn could be performed within the ability ole unit, the TM would not change design basis ec,nipment malfunctions. The H2O2 monitoring equipment is used post sacident to identify hydrogen concentration in the Containment and Drywell. This TM will

prevent one failure mode for the system that resulted from the tap change of the 120V distribution i

5

Attachment 2 .

to U 602:36 ;

, Page 129 of142 P

i transformer, There are no accident scenarios that ere initiated from failure of these panels.

Therefore, this change would not affect design basis accidents. This TM is contained within existing circuits of the H202 monitoring system, maintains the system functionality and would not -

L create new design basis accidents or malfunctions. This TM maintains the operation of the H202  !

l monitoring equipment.- Maintaining operability would not reduce any margin of safety. .

TEMPORARY POWER TO TEMPERATURE CONTROLLER RACK IN PANEL IH13-P868

. Document Evah'ated: TM 97-038- LogNumber: 97-140 -

i

This Temporary Modification (TM) addresses one of the overvoltage concerns raised during the review of the circuits fed from the 120V distribution panel which are having their transformer taps i changed to provide a 5% increase. The power supply for the temperature controller rack in the i

electronic control module portion of the H2O2 monitoring system has limitations on voltage input. The power supply provides operating voltage to tne temperature controller rack which controls the heat tracing for various sampling lines in the system. The TM uses the control circuit of the air compressor for the H2O2 sampling panel (ICM01SB) to energize the power supply for the temperature controller rack in panel lH13-P868. This mainta:ns the divisional power

_ requirements of the equipment as well as the description of the cc '.nment_ monitoring svstem in ,

' the USAR. However, schematic diagrams, listed in USAR Table 1.7-1, for the system are impacted by the revised circuit wiring. This TM will be installed through operating Cycle-7 until permanent degraded voltage modifications are made.

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. The H2O2 monitoring equipment is used poet accident to identify hydrogen concentration in the Containment and Drywell. 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.

Therefore, this change would not affect design basis accidents. This TM is contained within existing circuits of the H202 monitoring system, maintains the system functionality and would not create new design basis accidents or aulfunctions. This TM maintains the operation of the H202 monitoring equipment. Maintaining operability would not reduce any margin of safety.

INSTALL CARPET IN THE RADWASTE OPERATIONS CENTER (ROC)

Document Evaluated: ECN 30205 Log Number: 97-142 This ECN results in installing carpeting in the Radwaste Opa ations Center (ROC). The carpet installed is similar to that installed in the main control room (MCR) under ECN 29418 (safety evaluation Log Number 95-092). The carpet installation is intended to improve aesthetics, acoustics and the general working environment in the ROC. Floor covering descriptions provided in USAR Appendix E, Section 3.1,2.2.8 are affected by this change. Condition Report (CR) 1-97-05-163 identified that this section should have been revised to add carpet description as a result of ECN 29418. The carpet information added by that change was applicable to both the

m _ -

Atti.chment 2 '

to U 602836

. Page 130 of 142 MCR and ROC, The description of Fire Zone R-10, which includes the ROC as described la USAR Appendix E, Section 3.8.1.14 is also affected by the VCN. In addition, Appendix E, Section 4 (C3mpliance with '3TP APCSB 9.5-1) will be revised by the ECN to include the ROC L

in the dkcussion.

The ROC is the control conter for monitoring and controlling the solid and liquid radwaste systems. It is needed for power generation purposes only and has no safety functions. As stated in USAR Appendix E, "the radwaste building houses no safety-related equipment." In addition, calculation IP-M0177, Rev. 4 Volume C, determined that there is no change in area fire severity rating as n ' wit of thiese. Therefore,- this change would not impact design basis equipment malfunction . Changing floor covering types do not impact design basis accidents. Since the

- radwaste building is separated from safety-related areas by 3-hour barriers, this change would not create new .esign basis accidents or malfunctions. The radwaste facility and carpeting do not h?fect any acceptance limits; therefore, margin of afety would not be impacted by the change.

TEMPORARY MODIFICATION FOR THE DIESEL GENERnTOR VENTILATION (VD)

SYSTEM Docement Evaluated: TMs97-048,97-046 and 97-045 Log Number: 97-145 Temporary Modification (TM)97-048 raised the setpoint for the Diesel Generator ventilation system supply air heater controller for the Division 1, 2, and 3 Diesel Generator (DG) rooms.

TMs97-046 and 97-045 increase the Division 1 and 2 DG room supply fan damper temperature controllers from a minimum temperature of 70'F to between 35 and 90 F when the DGs are in operation. These changes allow the diesel generators to continue to meet the design basis system.

Changing the temperature of the DG rooms effectively reduces the horsepower requirements for

. the VD fans and therefore reduces the DG load. This change to Division 1 and 2 ensures bus voltages will be maintained.

This change will not change design basis equipment malfunctions since the change ensures that bus loading for Divisions 1 and 2 is not exceeded and that equipment located in all three division rooms will be operated within the temperature bands previously evaluated. This change would not affect design basis accidents since the systems will operate as previously evaluated. This

.would also not create new accidents or equipment malfunctions for the same reasons. This change will not affect the margin of safety since equipment will be operated within system requirements and with parameters established in Design Criteria DC-VD-01-CP.

OPERATE DURING CYCLE-7 WITH AREAS OF UNCOATED CARBON STEEL IN CONTAINMENT Document Evaluated: CR l-97-05-014 LogNumber: 97-146 US AR Section 6.1.2.1 states that exposed carbon steel (CS) surfaces inside containment are to be coated with an inorganic zinc primer which has been qualified in accordance with ANSI Standards N101.2, " Protective Coatings (Paints) For Light-Water-Cooled Nuclear Reactor Containment

Attachment 2 to U 602836

_ Page 131 of142 Facilities," N101.4," Quality Assurance for Protective Coatings Applied to Nuclear Facilities,"

Regulatory Guide 1.54, " Quality Assurance Requirements for Protective Coatings Applied to -

p z Water Cooled Nuclear Power Plants," and ANSI N512, " Protective Coatings (Paints) For the -

i Nuclear Industry." Several areas including the containment liner were also coated with epoxy _

l coatings. USAR Section 12.1.2.2.3 states that "(For ease of decontamination), walls and floors

- are coated to a smcoth finish in plant ueas where contamination is possible." In response to

_ Emergency Core Cooling System (ECCS) suction strainer clogging concerns identified during j RF-6, areas in containment have been stripped, and most of the stripped areas recoated. 1 However, not all stripped areas have been recoated. While it is the intent to recoat the bare areas -

l_ during plant operation, some areas may remain uncoated throughout operating Cyde-7. The i

evaluation is to determine that an unreviewed safety question does not exist while in this

configuration. The four basic issues to review are

(1) additional debris due to corrosion product

from the exposed CS in Containment, (2) integrity of the containment liner paint, (3) hydrogen

production following a LOCA, and (4) the ability to decontaminate the uncoated surfaces.

' The containment liner and the ECCS suction strainers are considered important to safety _

equipment. The evaluation determined that the amount of rust buildup on exposed areas for one

cycle of operation to be insufficient to impact st ainer operability or containmerit liner integrity.

In addition, the areas left uncoated would not present a significant decontamination problem since i the total area is small in comparison to the total contaimnent surface area. Therefore, design basis

, equipment malfunctions would not be affected by this condition. Since operability of the j' equipment will be maintained, no new equipment malfunctions would be expected. The amount l' of hydrogen generated from this exposed areas would not impact the 4% flammability limits as addressed in USAR Section 6.2.5.1.3.2. In addition, liner integrity would not be affected by the

!- - nominal corrosion expected in one cycle of operation. Therefore, containment barriers and the

[ margin of safety would not be affected, i

j DE-ENERGIZE MCR INDICATING LIGHTS (1VX03YA) -

i

- Document Evaluated:- TM 97-052 LogNumber: 97-149

This Temporary Modification (TM) de-energizes the Main Control Room (MCR) position i- indicating lights for the IVX03YA damper by lifting the AC line feed from the control circuit in i- . MCC 1 AP72E. The remainder of the control circuit wnains in operation and the Switchgear i- Heat Removal (VX) System damper will continue to function to support operations of fan

[ IVX03CA. This fan is required for chiller 1VXO6CA operability. The purpose of this change is 1 to prevent the possibility of remote damage to the circuit (such as a MCR fire) which would make

the damper inoperable. This TM will remain in place until a permanent design change is installed

- to resolve this concert. This TM impacts a schematic drawing which is part of the USAR per Table 1.7-1. The VX operating procedure CPS procedure 3412.01 will will be changed to check y - damper position at the local indicator position at IPL6."A on the 781-foot level of the Auxiliary i Building.

4 This TM eliminates the possibility of equipment malfunction by removing the possibility of

damaging the control circuits fuse. Since this TM de-energizes the indicating light not required for operation of the damper, there is no change to design basis equipment malfunctions. This TM l

i-

Attachment 2 to U402836 Page 132 of142 is intended to ensure IVXO3YA damper operability and the Division I VX system and equipment -

it cools.- As such, this TM insures that the Safe Shutdown Analysi.3 (SSA)in the USAR will continue to be met. Thus, this change wou!.} not negatively impact design basis accidents. Since the damper will continue to operate as designed, the TM would not cause new accidents or

_ equipment malfunctions. This TM supports the SSA by ensuring that VX is available per design;

therefore, the margin of safety would not be reduced.

j - REVISION TO USAR FOR MISCELLANEOUS DRAIN SYSTEM (DV) CHANGES Document Evaluated: ECN 28876 LogNumber: 97-150 USAR Figures 3.6-1, Sheet 23 and 10.4-9, Sheet 2 are being revised to indicate the normal positi an of IDV036A from closed to open and to add two new drain valves, IDV036C ami 1DV036D, in 2-inch line, IDV30BA. This change is due to the poor performance of the IDV036A valve. It has been decided to abandon this valve and add two new vr.lves downstream to provide drain isolation capability. These valves provide tube side drr.in isolation on the 6A high pressure feedwater heater.

The new valves meet the pressure and temperature requirements for the system. The leak

_ perf ormance of these new valves is expected to improve based on extensive industry usage in I similar applications. The small weight increase made by the addition of these valves is well within the acceptable allowance of the associated piping. Therefore, this change would not impact design basis equipment malfunctions. Steam line breaks evaluated in the USAR are for major line breaks. Therefore, a break in the 2-inch line is bounded by a break in one of the large feedwater lines. Since these valves have extensive industry history of proven performance, and feedwater -

line breaks are already evaluated in the USAR, this change would not introduce new accidents or malfunctions, These valves are not safety-related nor affect any acceptance limits, therefore the margin of safety would not be reduced.

- DIESEL GENER.ATOR (DG) OPERABILITY DETERMINATION FOR DEGRADED CONDITIONS Document Evaluated: 1-97-06-062-OD Log Number: 97-156 Operability Determination (OD) 7-97-06-062-OD was prepared to address two diesel generator (DG) degraded conditions. The first condition evaluated was the lack of a 1/16-inch hole drilled on the poppet of the 30 psig relief check valves installed in the discharge line for each DG lube oil pump. A hole is drilled in the 30 psig check valve to relieve pressure trapped between the relief valve and the circulating lube oil pump check valve to allow for a low pressure alarm in event of a pump failure. Since none of the diesel generators at CPS have a pressure switch located between these check valves, this deviation from the vendor's recommended design will not adversely impact the function of any DG sisans. The second comlition evaluated was an inaccurate 4

-statement in which USAR Section 9.5.7 states that all three DGs are equipped with pressure switches to monitor operation of the circulating lube oil pump / motor set and turbo soakback pump / motor set. Divisions I and 11 have these features; however, Division III only has an' alarm for the circulating tube oil system. These ditierences are due to differences in designer

)

Attachment 2 to U 602836 Page 133 of142 considerations for Division I and II and those for Division III. USAR Figure 9.5-5 has been updated to redect the present configuration of the DG lubrication system. The USAR contains the original schematic of the DG lubrication system, which does not show the automatic pre-lubrication system installed in 1986 under modification DO-013. The updating of this figure is

. consistent with the description in US AR Section 9.5.7.

For the first condition, sight glasses were installed in the circulating lube oil line as part of modification DG-013 to provide positive indication that che oil galleries are full. This level is checked and logged by the area operator each shift. For the second condition, the area operator

- is to verify that oil is draining from the turbocharger gear train to the engine sump prior to any routine DG start. This verifies proper turbo soakback pump operation. These conditions would not affect the protective and supervisory functions for the DGs as provided in USAR Tables 8.3-4 and 8.3 18. Therefore, these conditions would not impact design basis equipment malfunctions. <

The failure of a DG is not an initiating event for accidents addressed in USAR Chapter 15. In addition, USAR Sections 9.5.7 and 8.3.1 state that failure of any one DG component will not result in the loss oflubricating oil supply to more than one DG. Therefore, two DGr temain available during a design basis accident. The failure modes in Section 9.5.7 and Table 9.5-11 are not affected by this change. The turbo soakback pump and the circulating oil pump are still available if the primary pump is lost. As such, this would not initiate new accidents or equipment malfunctions. The only critical DG lubrication parameter defined in the Technical Specificatior.

• Bases is total oil capacity, the above conditions do not affect the oil capacity and therefore do not reduce the margin of safety.

REVISION TO DIESEL GENERATOR ANNUNCIATION SYSTEMS Document Evaluated: CR l-97-05-167 LogNumber: 97-157 In preparation for the NRC electrical distribution system functional inspection (EDSFI), a review of the diesel generator (DG) annunciation information in the USAR was performed. The review identified some discrepancies in the description to the plant configuration and the design

- documents. This revision corrects the discrepancies found so the information contained in the USAR accurately reflects the as-designed and analyzed configuration of the DG annunciation

. system. These r.hanges affect USAR Section 8.3.1.1.12 and Tables 8.3-4 and 8.3-18.. It was determined that these design changes had been made in the past and not incorporated into the USAR. The position names of the control switch and additional alarms of the DG system have been revised. Operation of the DG system is not altered.

DG malfunctions of the fuel oil system, cooling water system, starting air system and lubrication system were reviewed. This change does not affect these failures nor does it affect the operation of the DG system. As such, equipment malfunctions would be unaffected. Accidents evaluated involve operator actions to verify proper switching and loading of the emergency DGs or verifying the DGs have started and are in st andby following an accident. The changes have no effect on the ability of the DG system to function in support of these accidents. These changes are minor and have no impact on the safety function of the system. Therefore, this would not cause new accidents or equipment failures. This change does not impact the Technical Specification Bases or the margin of safety since DG operation is the same.

Attaciuneet 2

, to U 602836 Page 134 of 142 L MINOR SERVICE WATER (WS) PIPING CHANGES Document Evaluated: ECN 28161,28788, et. al. LogNumber: 97-159 Between July 1994 and February 1996, several design changes were made to the Plant Service Water (WS) System requiring incorporation into the USAR. These changes were minor and were j evaluated under the screening process for safety evaluations and a full safety evaluation not

completed. The safety evaluation process was changed in mid-1997 to remove some of the exclusion criteria for USAR changes without a full safety evaluation. Previously sub nitted USAR

changes were reviewed to ensure they met the new criteria.

The changes consisted ofinstallation of a split-type mechanical seal on the service water pump lWS0lPA and a %-inch drain line and valve (ECN 28161). Two mechanical seal drain lines on this pump had a common connection 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 service water pumps lW50lPB and lWS0lPC (ECNs

! 28789 and 28780). Two high point vents and three drains were added to plant service water lines j- feeding component cooling water system heat exchangers 1CC01 AA and ICC01 AB (ECNs F 28869,29253 and 29607) Finally, an air volume booster was added to the actuator on control j_ - valve IWS079A which controls service water flow through non-essential plant chiller 0 WOO 2CA i (ECN 28980).

All changes affect piping shown in USAR Figure 9.2-1. 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 components. If a functional failure of a new component

'; occurred, that failure will not impact any safety related system from performing its function. A 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 _ As such, these changes would not impact equipment malfunctions. The WS system is not assumed to function following'an accident. USAR Section 15.2.10 describes a complete loss ofinstrument air. The change performed to add the air volume booster to a valve actuator does not impact this analysis or conclusion of patulated design basis accidents.' Based on the minor scope of the changes described, this would not contribute to new design basis accidents of equipraent malfunctions.

The WS and Instrument Air (IA) Systems contain no acceptance limits within the CPS Technical Specifications. Therefore, this enange will not decrease the masgin of sarety.

This change has not been fully implemented as of the cutoff date for USAR Revision 7. All ECNs -

with the exception of ECN 28980 have been implemented. An updated status of this change will be reported in Revision 8 to the USAR.

Attachment 2 to U 602836

Page 135 of142 REMOVAL OF OFFGAS (OG) SYSTEM FLOWMETERS -

Document Evaluated: ECN 29120 LogNumber: 97-162

- This ECN results in the removal of three flowmeters (IN66-R043 A, R043B and R045 A) from the

. Offgas (OG) System. These flowmeters allowed maintenance to be performed on the non-j operating train of the OG system while the plant is on-line. Taps are located in various points

. along Ae process piping for air lines to be installed and purge air used to remove any hydrogen which may accumulate during on-line maintenance. Each tap has an isolation valve c..d an installed flowmeter to provide positive indication of purge air flow. These taps and flowmeters ,

provide no function during normal system operation. Since the isolation valve is normally closed,

- the flowmeters are not a part of the normal system flow path. The ECN removes the flowmeters for ALARA purposes and to improve the work process. If on-line maintenance is required in the

future, temporary flowmeters will be utilized.- This innpacts USAR Figure 3.6-1, USAR Table 11.3-5 contains possible equipment malfunctions of the OG system and design

features to mitigate these failures. Removal of these flow meters, used for maintenance purposes, l' does not impact these equipment malfunctions. USAR Section 15.7.1 describes and evaluated the j~

worst possible postulated leak from the OG system. Since the meters are normally isolated from the gas pressure boundary, removal would not impact design basis accidents or create new accidents or malfunctions. The OG system pretreatment radiation levels are controlled by the Technical Specifications (TS). Since this change does not hupact the operation of the OG system, the TS and Bases are unaffected, therefore there is no mduction in the margin of safety _

SPECIALIST TRAINING DESCRIPTION Document Evaluated: USAR Section 13.2.1.1.1.G LogNumber: 97-164

- This change adds the words, "and supervisors," to three job classifications to ensure that USAR Section 13.2.1.1.1.6 is in agreement with NUREG-0737, " Clarification of TMI Action Plan Requirements," wording. USAR Section 13.2.1.1.1.0 provides a description of" specialist training" provided to I&C Technicians, Chemistry Technicians and Radiation Protection Technicians. This section states that these job classifications receive Core Damage Mitigation Training. The statement in the USAR is intended to assure compliance with NUREG-0737, Task II.B.4, " Training for Mitigating Core Damage," which states that managers and technicians in I&C, Health Physics and Chemistry Departments shall rece've training commensurate with their responsibilities. Since this change only affects a description of a training program, and training

~

remains the same, there are no changes to equipment malfunctions or accidents. As equipment operation is unchanFe.i and appropriate personnel receive Core Damage Mitigation Training, no new accidents or equipment malfunctions are created. Also, this change would not decrease any margin of safMy.

Attachment 2 -

to U 602836 --

Page 136 of142 ABANDON IN-PLACE HEAT TRACING Document F, valuated: ECN 29360- LogNumber: 97-165 ECN 2936u results in abandoning in-place lEat tracing from the radwaste equipment drain processing system (WE) demineralizer and pH adjustment skid piping and all piping downstream which no longer requires heat tracing The heat tracing will be removed from USAR Figures 11.2-2, Sheets 6,14,18 and 29. The heat tracing is no longer needed since the caustic concentration levels have been reduced such that solidification of the caustic will not occur at ambient temperature. .This change is based on historically low demands for caustic at Clinton

' Power Station since the WE demineralizers use non-regenerative resins and the radwaste tanks are rarely out of specification regarding pH levels. This change does not affect radwaste system pressure boundaries, system flows and pressures, or operation of the system. Therefore, the change does not impact design basis equipment malfunction or create new malfunctions. This change only affects non-safety related equipment, not credited in any design basis accident.- The heat tracing does not affect any acceptance limits nor _does it support safety-related equipment.

Therefore, this change would not impact protective barriers or the margin of safety.

DIESEL GENERATOR (DG) STARTING AIR SYSTEM Document Evaluated: ECN 30250 Log Number: 97-166 The licensing and design basis of the diesel generator (DG) Starting Air System (SAS) has been revised. This revises USAR Section 9.5.6 and Technical Specification (TS) Bases B3.8.3 to clearly stats that the air storage volume is sized such that the system is capable of starting the respective diesel engine five successive times without recharging. The rated air capacity is also being added to these sections. The rated air capacities are 93 cubic feet at 250 psig for the Divisions I and II DGs and 64 cubic feet at 240 psig for the Division III DG. Also, the number of successive starts has been reduced to three when pressure in the tank is below its rated capacity but noove the mirJunum allowable pressure. This change will continue to satisfy the multiple start capability required by the Clinton Power Station Technical Specifications (i.e.,200 psig).

The DG SAS has sufficient capacity to ensure reliable multiple starts of the DG at the Technical Specification limit without recharging the associated air receiver. Therefore, the change would not impact design basis equipment malfunctions and not redace any margin of safety. The capacity of the SAS is sufficient to ensure reliable starting of the DG during a LOOP /LOCA event as well as multiple starts as require ( by the Technical Specifications. The failure of a DG is not an initiating event for any accidents addressed in USAR Chapter 15. The consequences of a single failure of a DG is boundM by the failure analysis in USAR Chapter 6 and the accident analysis in USAR Chapter 15. Therefore, design basis accidents would not be affected or created by this change. Since each DG has two independent and redundant SASS, a failure of one SAS would not cause a DG failure.

l

Atts:hment 2 to U 602836 Page 137 of142 .,

CHANGE TO ROOM FUNCTIONS TO USAR FIGURE E Document Evaluated: USAR rigure 1,2 5 Log Number: 97-167 USAR Figure 1.2-5 is revised to correct the names of the Cold Chemistry Lab, the Chemical Storage Area and the Mens and Ladies' Rooms on the 737-foot elevation of the Control Building to the actual uses of these areas: Office Area, Chemistry Office and Storage. The room names listed on the drawing no longer desc:ibe the actual function cf the room. Changing the names of these rooms is primarily administrative. The change in the storage room is contre!!ed by Radiation Protection (RP) controlling access to the rooms; Since this change is primarily administrative, no equipment is operated differently, sym and component are also unaffected.

As such, this would not impact design basis equipment m .lfunctions, accidents or margin of safety.

USAR CHANGE PACKAGE 7-301 AND CPS PROCEDURE 3004.01 PAC 0296-97 Document Evaluated: USAR Change Package 7-301 and LogNumber: 97-169 CPS Procedure 3004.01, Turbine Startup and Generator Synchronization

' USAR revision 7-249, Safety Evaluation (SE) Log Number 97-102, states in Section 8.3.1.1.2 that during cycle 7 operations, the Emergency Reserve Auxiliary Transformer (ERAT) will be the primary source of offsite power for the Class IE AC power distribution system while the plant is -

in a startup or shutdown conditions. This USAR change implements the description of the physical changes made to the Reserve auxiliary Transformer (RAT) and ERAT. Changes to the RAT and ERAT transfc,rmer tap settings were accomplished by ECNs 30102 and 30169 (SE Log

- Numbers97-102 and 97-126 respectively). Due to low grid voltage experienced on the 345KV distribution system, Illinois Power (IP) has determined (Reference IP Letter U-602804 to NRC dated 8/1/97), that the more robust transformer should be the preferred source durig startup and shutdown conditions. This is substantiated by.various discussions in USAR Sections 8.1.5.1, 8.2.1.1, 8.2.1.2, 8.2.2.1, 8.3.1.2.1, and 8.3.1.1.2 which treat the RAT and ERAT as normal and alternate sources, each capable of supplying the required IE loads.- These USAR and CPS Procedure 3004.01 changes to incorporate a description of the physical _ changes and/or the changes in operating methods due to plant modifications previously evaluated does not impact the probability of failures of equipment important to safety, and do not introduce any new failure modes. These changes document previously evaluated changes and as such, do not affect the

_ probability or possibility of new or previously evaluated accidents. The physical plant changes were found not to reduce the margin of safety and as such, these USAR and procedure changes do not reduce the margin of safety.

DELETE SGTS HUMIDITY SENSORS '

Document Evaluated: USAR Figure 3.6-1 Sheet 93, et. al. LogNumber: 97-171 This change removes humidity sensors OUE-PR053 and OUE-PR054 from USAR Figures in USAR Sections 3.6,6.5 and 9.4. The instruments have been abandoned in-place. These

Anaciunent 2 to U402836 Page 13: of142 Instruments have no operational or performance requirements that are important to safety, in addition, this change deleted the input of these Instruments to channels 2, 3,4, and 5 of instruments OUlX PR051 and OUT PR055. The information provided by these instrumend are stack flow information for the Standby Gas Treatment System (SGTS) and HVAC systems, mmurirg and indicating SOfS dewpoint and temperature and HVAC dewpoint and temperature.

Chennel I still provides SGTS flow inte and Channel 6 provides HVAC flow rate. This j ,_

information is not used or required for any CPS calculations, procedures or processes.

! Failure modes of these instruments have not changed. These instruments are not safety related

equipment and do not afect important to safety equipment. The USAR accident analysis does i not take credit for these instmments. Therefore, heign basis accidents would not be afected by

] this change. Abandoning these instruments would not reduce the margin of safety.

]. CHANGE FP VALVE CONFIGURATION Docur..t Evaluated: ECN 29234 LogNumber: 97174 l ECN 29234 revised USAR Figure 9.5 1, Sheet 2 to reflect the outdoor fire protection (FP)

system isolation valves 0FP066 rad 0FP139 as normally closed, check valves 0FP067 and 0FP140

} as removed, and installed a blank flange at the point of building entry. The isolation valves were i

locked shut and blank flanges installed in place of the check valves to prevent the vet pipe sprinkler syrteir in the normally unmanned buildings (located outside the Protected Area) from

, freering during the winter mombs and to prevent possible leakage. Pipiy downstream of the isolation valves was not safety or quality related.- The installation of the blank flar ges in the two

buildings protect the integrity of the FP system by ensuiing that a leak in the downstream piping

! would not affect the FP system piping.

l Thr.wo buildings house no important to safen equipment. No new failures have been created and the changes have no effect on fire initiating events and will not change design basis accident 4

analysis. Since this change is confined to non-safety related portion of the FP system, design basis

. _ equipment malfhnctions would not be affected. This change has no effect on physical parameters 6 _I - which support any margin of safety as defined in the Technical Specification (TS) Bases.

RELOCATE AND CORRECT DESIGNATION OF VALVE STEM PACKING LEAK-OFF LINES

Document Evaluated

ECNs 28637 and 28990 LogNumber: 97178

! These ECNs reloc'.ted the Rewtor Core Isolation Cooling (RCIC) stem packing leak-offline on IE51F610 from the side of the stem case to the bottom. This change will improve drainage of i condensate which previously accumulated in the bottom of the stem case promoting stem j corrosion / pitting. The ECNs allowed the use of stainless steel tubing and fittings in the leak-off

[ line. The change also involved dri' ling two tapered dowel holes in the bonnet to realign the valve L linkage. The ECNs also corrected the turbine trip and throttle valve (IE51 C003) and turbine

governing valve (IE51F610) incorrect designations as valve packing leak-offlines. Lastly, the i

l l

=

Attachment 2 to U402836 Page 139 of 142 schematic for this system is being updated to reflect the turbine goverp.ing valve stem packing leak offline.

These changes did not affect the pressure-retaining capability of the bonnet or leak-offline. Also evaluated were seismic requirements due to the increase:1 weight. These aspects were determined to be unaffected by the change; therefore, this change does not impact design basis equipment l

malftnctions. USAR Section 5.4.6.1 (4) states that RCIC can be used in conjunction with High Preamre Core 3ptay (HPCS) for a design basis control rod drop accident. Moving the leakage line decreases the likelihood of a sticking governing vCv stem and thus assures the availability of RCIC for mitigating the consequences of this event. Use of stainless steel is acceptable as the material meets the system design requirements. The rerouting of the leakoffline will not change equipment malfunctions or decrease the margin of safety since reliability is assured by this change.

REVISE CONDENSATE POLISHING (CP) SYSTEM VALVE POSITION Document Evaluated: USAR Figure 10.4-4, Cheet 6 LogNumber: 97180 US AR Figure 10.4-4 Sheet 6 was changed to show the low conductivity, low crud (LCLC) tank drain pot valve ICPLCV4A as normally open. This valve is normally open to allow the drain pot which catches effluent ftom various relief valves to drain to the high conductivity, high crud

. sump. This valve operates opposite the 1 CLC tank chemical waste divert valve, ICPLCV4, which is normally closed. The divert valve la designcd to open when high conductivity is sensed in the liquid flowing from the regeneration tanks to the ICP14T tank. This diverts the high conductivity liquid to the high conductivity sump. To prevent this liquid from entering the drain pot, the drain pot valve closes.

This valve is not safety related and does not interface with, or impact any, safety related equipment. This changes the USAR to reflect actual operation of this system. As such, this valve lineup change wauld not impact design basis equipment failurcs, The only US AR accident potentially impacted would be USAR Section 15.7.3, " Postulated Radioactive Release Due to Liquid Radwaste Tank Failure." This valve does not directly interface with any radwaste system tank, nor does the change impact any of the assumptions made in the analysis. The opeation of the Condensate Polish ng (CP) System is not governed by the Technical Specifications or Bases; therefore, this does not impact any margin of safety, REVISE BREATHING AIR (RA) SYSTEM VALVE POSIflONS

- Document Evaluated: USAR Figure 9.3 3, Sheet 8 Log Number: 97-182 USAR Figure 9.3 3, Sheet 8, was revised to show valves IRA 001 A/B, IRA 002A/B and IRA 003B as normally closed. These valves are isolation valves in the lines from the breathing air fill station outside the Diesel Generator (DG) Building and are normally closed. Valves

- 1RA003 A/h are the isolation valves between the storage bottles and the Main Control Room (MCR) ring header. During normal operation, one (A or B) valve la oper and the other is closed.

This change will eliminate potential confusion regarding the normal position of the valves.

Anaciunent 2 to U4028%

Page 140 of142 USAR Section 9.3.1.3.a, states that the RA system has no safety related ibnction. Failure of the RA system will not compromise nuclear safety nor prevent the safe shutdown of the reactor. This change does not afect the piping classi6 cation of RA piping. As such, this would not afect design basis equipment malfbnctions. The RA system is not depended upon to prevut or mitigate design basis accidente. This clari6 cation will not change equipment operation. Therefore, new accidents or malfbnctions would not be created. The RA system is not Technical Speci6 cation related; therefore, this change would not impact the margin of safety.

REVISE DAMPER SETTINGS IN MACHINE SHOP VENTILATION SYSTEM Document Evaluated: ECN 29310 Log Number: 97-183 This change allows opening damper OVJ102Y, located in the Radiological Controlled Area (RCA) boundary, and closing of dampers OVJ91Y and OVJ96Y located outside the RCA boundary, in the Machine Shop (VJ) HVAC system. Because the RCA boundary in the Machine Shop area was extended, some of the machinery exhaust ducts outside the RCA are required to be closed to increase the air flow to the remaining ducts to ensure air flows from clean to contaminated areas. Total aldlow into and out of the Machine Shop (5000 CFM) is unasected.

The ECN also speci6ed installing locks on the three shut dampers in order to ensure positive radiological control.

The VJ system has no safety basis according to USAR Section 9.4.9.1.1. Therefore, this change does not impact design basis equipment failures. This change does not impact design basis accidents since the VJ system is not relied upon to mitigate the consequences of an accidot. The ch .nges to the VJ system do not reduce any margin of safety. -

P2 VISE DESIGN DOCUVENTS TO INCLUDE OMITTED PRESSURE GAUGES AND VALVES Document Evaluated: ECN 2%55 Log Number: 97-194 This ECN added pressure gauges and their associated bit-of valves located on the Automatic Depressurization System (ADS) backup air bottle racks. These gauges and valves were originally

, provided with the vendor supplied ADS backup air supply skids, however, they were not reflected

- in the design drawings. Two gauges and associated valves are installed on each of the two divisions of ADS backup air bottle racks, a high pressure gauge on each of the bottle rack headers

- and a low pressure gauge on each of the panels. This changes USAR Figure 9.3 2, Sheet 5.

The ADS provides a method for operators to manually depressurize the vessel during emergency conditions. The pressure gauges and associated shut-oKvalves are considered passive safety-related devices since their only safety function is to maintain pressure integrity. The pressure gauges and valves are designed to applicable ANSI standards and seismically qualified. Since these components are built to the same standards as the remaining ADS backup air systems, this change will not increase the probability of equipment malfunctions important to safety. The backup air system is normally isolated from ADS. Failure of a pressure gauge or valve pressure boundary will result in isolation of only one of the divisions of the backup ' air system. Therefore, 1__- -- -

- l ' . -

i Attachment 2  !

• to U 602:36 l Pase 141 of142 i

!l I

this change would not impact the probability of design basis accidents. Failure of these valves and 4

gauges could generate missiles. However, since the components are built to the same standards  ;

4 u the remainder of the backup air system, the change would not introduce new accidents not  !

already evaluated or new equipment malfunctions. The addition of these gauges would not affect  !

the design basis fbnction of the ADS. The ability of the Energy Core Cooling System (ECCS)  !

systems to provide emergency cooling is not compromised by this change. Therefore, the margin of safety would not be impacted. .

, USAR CHANGE FOR CORNBELT FS AND TERRA INTERNATIONAL ,

1  !

i Document Evaluated: US AR Sections 2;1, 2.2 Log Number: 97 201

- USAR Chapter 2 was revised to change the name of an industry within five miles from Clinton  !

Power Station and to include a description of the hazards from additional storage of ammonia. ,

Shields Soll Service was renamed Terra Intemational and added additional storage of ammonia.

s Cornbelt FS has also added propane storage to its facility. Terra Intemational previously had one  :

40-ton ammonia tank. Another 40 ton tank has since been added. Cornbelt FS has added l propane storage. The amount stored is less than 1,000,000 pounds. The calculated safety level 3 for propane was determined to be 13,240,000 pounds. Therefore, the 1,000,000 pounds would not impact plant equipment or create design basis equipment malfunctions. This evaluation  !

determined that the possibility of any single incapacitating event resulting from one or both tanks of ammonia rupturing is sufficiently low enough not to be considered a design basis event. Since L this change does not impact any acceptance limits, the margin of safety is not reduced.

! DISCONNECT AND CAP INSTRUMENT AIR BRANCH LINE TO RCIC TESTABLE L _ CHECK VALVE i

Document Evaluated

ECN 27735 LogNumber: 97 207

\' .

This change disconnected and capped a leaking 1-inch Instrument Air (IA) system branch line, IIA 82C, from its supply header (3 inch drywell IA ring header IIA 13E). This branch line is not safety related or seismically qualified. This line originally supplied IA to valve 1E51F066; i however, Modification RI-043 eliminated the need for IA tu the Reactor Core Isolation Cooling (RCIC) testable check valve, to this valve. The vsNe no longer needed air as it was converted to -

a simple check valve. This modification wa eported in USAR Revision 5, Safety Evaluation Log Number 93-0109. This change uphes USAR Figure 9.3-2, Sheet 12 to show the branch line disconnected from its supply header and capped off.

Since this line is no longer needed to support IE51F066 as a simple check valve and supplies no other instruments or equipment, this change does not impact design basis equipment malfunctions.

The line was acceptably evaluated for seismic qualification in an abandoned state. Since the j change dces not affect any imponant to safety functions, the change will not increase the

. - probability of an accident. Disconnecting and capping this unnecessary line would not introduce ,

l new malfunctions or accidents, nor reduce the margin of safety.

4 y

--m ,v. a w---N w , e ,ew y - rw- -,n, - - - - --,~-,en ---~c -~~ - - ~ ~ , - r,~- - - e,

Attachment 2 l 0 U 602836 l Page 142 of 142 CR IDENTIFIED TilAT US AR CilANGE WENT BEYOND BEING ADMINISTRATIVE IN NATURE ,

Document Evaluated: CR l 97-05-198/ECN 27959 LogNumber: 97 208 ECN 27959 changed the use of the Chemistry Lab, Cold Lab area to an office area. The chemistry hoods and controls were removed from the laboratory and the supply and return ducts previously connected to the lab hoods were disconnected. The changes affect USAR Figures 9.4 18, Sheet 3 and Figure 1.2 5. This change is related to Safety Evaluation Log Number 97-167 reported earlierin this revision.

i This change will not impact design basis equiprtant malfunctions as the Laboratory IIVAC (VL) system does not affect important safety equipment. Fire initiating events have been reduced because laboratory vek is no longer conducted in this areas. No new failures were identified and the changes do not reduce margin of safety.

I e

l

, - - - - ,- , w-