ML20199J514

CPSES 10CFR50.59 Evaluation Summary Rept 0007,for 960202- 970801

ML20199J514
Person / Time
Site:	Comanche Peak
Issue date:	02/01/1997
From:	TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
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ML20199J510	List: ML20199J504 ML20199J514 ML20199J539
References
NUDOCS 9802050333
Download: ML20199J514 (111)

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COMANCHE PEAK STEAM ELECTRIC STATION i

10CFR50.59 EVALUATION

SUMMARY

REPORT 0007 FEBRUARY 2, 1996 AUGUST 1, 1997 TEXAS UTILITIES ELECTRIC COMPANY f

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- to TXX 98021 Pa9e 2 of 110 COMANCHE PEAK UNITS 1 AND 2 10CTR50.59 EVALUATION

SUMMARY

REPORT 0006 TABLE OF CONTENTS i

This report contains a description and a summary of the following 10CFR50.59 Evaluations:

Eval No.

Revision LDCR No.

Eval No.

Revision LOCR No.-

i SE 90 041 Rev.

5 SA 96 108 SE 96 065 Rev.

O

~

SE 91 062 Rev.

8 SE 96 066 Rev.

1 SA 96 110 SE 93 109 Rev.

1 SA 96 80 SE 96 068 Rev.

O SA 96 134 s

SE 94 036 Rev.

O SE 97 001 Rev.

1 SA 97 1 SE 94 038 Rev.

O SA 94 150 SE 97 002 Rev.

O SA 9619 SE 97 003 Rev.

O SA 97 6 SE 94 043 Rev.

1 SA 94 90 SE 97 004 Rev.

O SA 97 5 SE 94 048 Rev.

3 SA 94 143 SE 97 006 Rev.

O SA 97 10 SE 94 095 Rev.

1 SA 94 153 SE 97 007 Rev.

O SE 95 018 Rev.

O SA 95 58 SE 97 009 Rev.

O SA 97 20 SE 95 033 Rev.

O SE 97 010 Rev.

O SA 97 25 Se 95 040 Rev.

1 SA 96 55 SE 97 011 Rev.

O SA 96 56 SE 97 012 Rev.

O SA 97 24 SE 95 041 Rev.

1 SA 95 72 SE 97 013 Rev.

0 SA 95 89 SE 97 016 Rev.

O SA 97 27 SA 96 92 SE 97 017 Rev.

O SA 97 28 SE 95 044 Rev.

O SA 95 101 SE 97 018 Rev.

O SA 97 31 SE 95 056 Rev.

1 SA 95 122 SE 97 019 Rev.

1 SE 96 006 Rev.

O SA 96 8 SE 97 020 Rev.

O SA 97 35 SE 96 020 Rev.

O SE 97 021 Rev.

O SA 95 106 SE 96 021 Rev.

O SE 97 022 Rev.

O SE 96 024 Rev.

O SE 97 024 Rev.

O SE 96 027 Rev.

O SA 96 44 SE 97 026 Rev.

O SE 96 029 Rev.

O SA 96 66 SE 97 027 Rev.

O SA 97 36 SE 96 032 Rev.

O SA 96 78 SE 97 028 Rev.

O SE 96 034 Rev.

1 SA 96 60 SE 97 032 Rev.

O SE 96 038 Rev.

O SA 96 ?4 SE 97 035

Rev, O

SA 96 131 SE 96 039 Rev.

O SE 97 036 Rev.

0 SA 97 47 SE 96 040 Rev.

O SA 96 76 SE 97 037 Rev.

O SA 97 49 SE 96 041 Rev.

O SA 96 75 SE 97 038 Rev.

O SA 97 51 SE 96 042 Rev.

O SE 97 039 Rev.

2 SA 97 7 SE 96 044 Rev.

O SA 96 83 SE 97 040 Rev.

O SE 96 045 Rev.

O SA 96 85 SE 97 041 Rev.

O SE 96 047 Rev.

O SA 96 88 SE 97 043 Rev.

O SE 96 048 Rev.

O SE 97 044 Rev.

1 SA 97 142 SE 96 049 Rev.

O SE 97 048 Rev.

O SA 97 77 SE 96 050 Rev.

0 SE 97 049 Rev.

O SE 96 053 Rev.

O SE 97 052 Rev.

O SA 96 34 SE 96 056.

Rev.

0 SA 96 100 SE 97 053 Rev.

O SE 96 057 Rev.

O SA 96 82 SE 97 054 Rev..

O SE 96 058 Rev.

O SE 97 056 Rev.- 0 SA 97 79 SE 96 059 Rev.

O SA 96 99 SE 97 057 Rev.

O SA 97 83 SE 96 060 Rev.

O SE 97 059 Rev.

O SA 97 85 SE 96 061 Rev.

O SA 96 105 SE 97 ^61 Rev.

O SE 96 062 Rev.

0-SE 97 065 Rev.

O SE 96 063 Rev.

O SA 96 107-SE 97 066 Rev.

O SA 97 99 SE 96 064_

Rev.

1 SE 97 067 Rev.

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Attaclaent 1 to TXX 98021 Pa9e 3 of 110 i

f Eval No.

Revision LDCR No.

SE.97 069 Rev.

O SA 97 104 l

SE 97 073 Rev.

0 SE 97 074 Rev.

O

.SE 97 076 Rev.

O SA 97 126 e

SE 97 077 Rev.

0 l

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SE 97 078

Rev, O

SA 97 133

~

SE 97 079 Rev.

O SA 97 124 SE 97 080 Rev.

0 SA 97 125 SE 97 081 Rev.

O f

SE 97 082 Rev.

O SA 97 117 SE 97 083 Rev.

0 l

SE 97 085 Rev.

0

+

2 SE 97 086-Rev.

O SA 97 50

-j SE 97 088 Rev.

0 SE 97 090 Rev.

O SA 97 160

.l SE 98 003 Rev.

O SA 97 9 i

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' to TXX 98021 Page 4 of 110 Evaluation Number SE 90 041 Revision 5 Unit: 1N2 Activity

Title:

Revision to the Post LOCA Hydrogen Generation Evaluation to Update the Evaluation with the latest Containment Temp / time Curve Description of Change (s):

The analysis described in the updated Final Safety Analysis Report was based on a Westinghouse Evaluation documentu in WPT-15218. This evaluation is updated by a new evaluation documented in WPT-15314. The new evaluullon accounts for the addit!on of 3125 lb. of Zircaloy cladding weight,9649 30. ft, of Zinc, the use of the Ickst enveloping LOCA Containment Temperature vs. time curve, and considers an extended fuel cycle.

Summary of Evaluation:

The possibility of increased radiation levels due to additional aluminum or zinc Inside Containment depends primarily on the possibility of parts or materials containing these elements being exposed to an Intense neutron flux during power operation, thereby becoming irradiated. It 18 not expected that such parts or materials could inadvertently be allowed to enter the Reactor Coolant system, thus passing through the core neutron flux region, while the Reactor is a power, Any potentialincrease in dose rates due to Al 28 would be of short duration after reactor shutdown and have insignificant radiologicalimpact. Even an unreasonable large quantity of irradiated zinc assumed to be released to the Containment would yield post accident dose rates which are negligible in comparison to those from the-fiss,an products postulated to be released per Regulatory Guide 1.4. Assuming that all of the Al and Zn in Containment dissowes into the aqueous phase, this would produce a very dilute solution of metal cations in the watar. A reactor would then be required to deposit the Al or Zn in the stainless steel grain boundarles (the chance of metallon undergoing the electrochemical reduction directly on a grain boundary are very small). Since the RCS is cool (below 200'F) and depressurized, the effect of embrittling would not be a problem. The temperature needed to " soften" the grain boundary is not present and the stress needed to propagate a crack is not 3 resent. The direct effect to the Containment atmospheric pressure will be a net increase of ess than 1.0% following LOCA; therefore, negligible.

.. to TXX-OS021 Page 5 of 110 Evaluation Numt,er SE. 91062

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Revision 8 Unit: 1X2 Activity

Title:

Radioactive material and radioactive waste handling and staging in areas outside the plant Description of Change (s):

Due to insufficient space inside the plant, designated areas outside the plant are required for radioactive material and radioactlU waste handling and staging. The fenced area east of the Fuol Building and areas in adjacerat to Warehouse C will be used for radioactive material and radioactive waste handling and staging.

Xevision 8: This revia!cn adds abrasive blasting of contaminated cornponents within suitable enclosures maintained under negative pressu.9 to the list of evaluatec activities.

Summary of Evaluation:

This evaluation considered normal operations in these areas airJ potential waste handling mishaps. It was determined that this activity does not involve an unreviewed safety question because the Impacts of the credible mishaps are enveloped by existing analyses that are within 10CFR100 limits.

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_ to TXX 98021 Page 6 of 110 '

Evaluation Number SE 93-109 Revision 1 Unit: 1N2 Activit; ' itle:

FSAR Chapter 11 revisions due to an extended fuel cycle (LDCR SA 93-126 and SA 96-080)

Description of Change (s):

Due to an extended fuel cycle, the fission product inventory in the reactor core will increase and result in changes to the normal operation radioactive effluents released from the plant.

Analyses have been performed to determine the potential result of these changes to offsite normal opeption radiation doses and radioactive concentrations in released effluent, waste tanks, and tu ste piocessing lines. Revision 0 of this safety evaluation considered the affect of assuming a apresentative reactor core design and Revision 1 of this safety evaluation considers a more bounding reactor core design.

Summary of Evaluation:

An extended fuel cycle willincrease the fission product inventory in the reactor core and result in changes to the normal operation radioactive gaseous and liquid effluents released from the plant. Although the analyses are not safety related, the potential offsite radiation doses will increase and, with conservatism, exceed 10CFR50, Appendix 1, Docket RM-50-2 design objectives, but remain within the 10CCR50, Appendix 1, design criteria. The commitment to RM 50-2 was met by the liquid waste pocessing system as described in the FSAR for a 12 month fuel cycle during construction, w,sich at that time, eliminated the requirement to perform a cost benefit analysis for the system. During operation, the criteria of 10CFR50, Appendix 1, apply and CPSES remains within these governing limits for the more bounding reactor core design. Accident analysis considerations. Including radiological consequences, are addressed in separate safety evaluations for each reactor core cycle configuration.

e to TXX-98021 Page 7 Cf 110 Evaluation Number SE-94 036 Revision 0 Unit: 1NN Activity

Title:

Installation of Bypass Test Instrument Prior to NRC Approval Description of Change (s):

Safety Evaluation SE 94-043 addresses the implementation of the Unit 1 Design Modification for Rcactor Protection System Testing In Bypass (i.e. Installation and operation). In order to fulfill in) requirements of 10CFR50.59, Nuclear Regulatory Commission approval is required prior to cperating the new design change because the pbnt Technical Specifications need to v rev; sed. CPSES willimplement the design modification (i.e., limited to installation)in future ottages but wii! only operate this plant modification after receipt of approval from the NRC.

Since this design modification will be installed, this safety evaluation is provided to evaluate the changes imposed on the plant faci'ities to ensure that there is no adverse impact to the existing plant design during operation of the plant prior to obtaining NRC approval. The installation of this design modification will provide the following plant changes: The Reactor Protection System Testing in Bypass Test instrument (BTl) introduces bypass testing circuitry for the Nuclear !nstrumentation Syttem power range reactor trip functions and the 7300 Process Proter.tlon System reactor tr!p functions and Engineered Safety Features functions. With the implementation of this modification, the likelihood of spurious reactor trips or safeguards actuation will be reduced, since a significant number of partiai trip conditions are eliminated.

No modification to coincident logic, trip setpoint, surveillance.equirements are associated with this change. Also, the BTI provides wiring from outputs of various cards into 50 pin connectors mounted on the 7300 & NIS cabinets. This will accommodate testing and trouble shooting act'vities without having to enter the cabinets. This modification will provide for an Automatic Test Equipment (ATE) to interface with the various card outputs via the 50 pin connectors mounted on the cabinet. When an ATE is developed, a safety evaluation will be prepared to address the interfacing of equipment via the 50 pin connectors. Since this modification is limited to the wiring only (i.e. without an ATE Interface) it is concluded that the ATE wiring outputs as evaluated in this safety evaluation do not introduce any unreviewed safety questions or require a technical specification change beyond what has already been identified as a part of this BTl modification, and can be used with standard test and troubleshooting equipment without prior NRC approval or further evaluation pursuant to 10CFR50.59.

Summary of Evaluation:

The installation of the Bypass Test Instrument (BTl)is to provide for future operation after NRC approval of license amendment request. If the BTI is inadvertently operated, there are alarms inside of the control room to alert the operatcts. Alarm Response Procedures are available for the operators to appropriately respord in a manner to ensure plant safety. The conclusion regarding an inadvertent operation of the BTI can be summarized as below (Reference SE 043): The implementation of this design modification does not result in a failure mode which increases the consequence of a malfunction of equipment important to safety. The proposed change does not alter the manner in which acceptance limits, limiting safety system set points, or limiting conditions for operation are determined. The change does not reduce the margin of safety.

1 to TXX-98021 Page 8 of 110 Evaluation Number SE-94-038 Revision 0 Unit: 1X2 Activity

Title:

Addition of 1E Fan Coll Units in Each UPS Room; Revision to FPR/FSAR Sections Ill/9.4 to Reflect changes Description of Change (s):

The design modification adds new indivW.:al safety related, Class 1E fan coil units to the Uninterruptible Power Supply (UPS) rNms with the fan coil units powered from the respective 480 volt 1E safeguard busses. Additionally in order to provide cooling of the cooling coils by the safety chilled water system, safety chilled water pumps are modified by replacement with larger size impellers. On completion of the above modification, the existing UPS chiller units are retained as redundant to the new fan coil units and placed in standby operation. The final safety analysis report (FSAR) section 9.4 is being revised to reflect the modification appropriately. Also the fire protection report (FPR) section lliis revised to add the new fan coil units to the fire safe shutdown equipment list.

Summary of Eva!uation:

The existing UPS room HVAC system consists of two 100 percent A/C chiller units which provide cooling for the four UPS Equipment and Distribution rooms. The UPS A/C units are common to both units at CPSES and function tc maintain the UPS Equipmant and Distribution rooms below the technical specification ambient temperature limit. This modification adds a safety related Class 1E fan coil unit to each UPS Equipment and Distribution room which in conjunction with the existing UPS A/C units provide redundant safety related cooling. The new fan coil units provide cooling to the UPS rooms utilizing cooling water from the safety chilled water system. The electrical power is supplied from the 480 voit Class 1E MCCs. The new fan coil units are controlled from a local panel mounted on the unit. The two existing UPS A/C units are maintained 100 percent to function as redundant safety related backup cooling. The modifications to all plant systems, components and structures are in accordance with all design and licensing basis requirements. Based on this evaluation, implementation of these activities do not involve an unreviewed safety question,

. to TXX-98021 Page 9 of 110 -

Evaluation Number SE-94-043 Revision i Unit: 1NN Activity

Title:

Unit 1 Design Modification for Reactor Protection System Testing in Bypass

)

Description of Change (s):

The Unit i Reactor Protection System Testing in Bypass design modification, introduces bypass testing circuitry for the Nuclear instrumentation System power range reactor trip functions and the 7300 Process Protection System reactor trip functions and Engineered Safety Feature functions. With the implementation of the modification, spurious reactor trip or safeguards actuation will be avoided since the partial trip conditions are eliminated. No modification to coincident logic, trip setpoint, or surveillance requirements are associatad with this change.

4 Summary of Evaluation:

The evaluation concludes that the implementation of the subject modification does not result in a failure mode which increases the consequence of a malfunction of equipment important to safety. Surveillance testing in bypass does not affect accident initiation sequences or response scenarlos as modeled in the safety analyses. The proposed change does not alter the manner in which acceptance limits, limiting safety system setpoints, or limiting conditions of operation are determined. The change does not reduce the margia of safety. Testing in bypass is expected to result in an overall improvement in safety by reducing unnecessary transients and challenges to the protective system by minimizing partial trip which may lead to inadvertent trips. Based upon the results of this evaluation, implementation of the proposed activity does not involve an unreviewed safety question.

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~ Attachment i 12 TXX-98021 Page 10 of 110 Evaluation Number SE.94 048 Revision 3 Unit: 1NN Activity

Title:

DM 93-061 INSTALLATION OF WOODWARD 701 GOVERNOR SYSTEM ON DIESEL GENERATORS Description of Change (s):

-This modification involves installation of a new control system for the emergency diesel governors such that slow start testing is possible as recommended by NUREG 1431 and GL 84-15. The maintenance and most of the surveillance starts will be done " slow" as opposed to the current " fast" starts. This slow start testing is expected to reduce engine wear and increase reliability. Existing EGA governors are becoming obsolete and manufacturer support is minimal. The EGA box, Motor Operated Potentiometer (MOP) and Resister Box of the existing governor system are left in place. A 701 Governor, Generator Load Sensor (GLS) and

- Magnetic Pick Up (MPU) Signal selector are installed in a new panel in the EDG room for the new governor system. The old components will be left as spares for the remaining EGA installations and to avoid requalification of the existing electrical panel. New speed sensors will be added to the EDG. A series of small holes will be added to the shaft in the area of the journal bearing area to accommodate activating the new speed sensors. A new EGB actuatcr will be installed.

I Summary of Evaluation:

The new digital governor systems offers new potential failure modes of Electromagnetic Interference (EMI) and common mode failure from software. However the EMI and common mode failure from software are evaluated as not probable occurrences. The installation accounts for minimizing the effects of EMI and a test was performed to confirm actual EMI levels are acceptable. The software is in widespread commercial service and has been Validated & verified by the sole source vendor (Cooper) for nuclear service. This modification enhances reliability of the EDG due to elimination of a significant number of fast starts which that cause excessive wear on engine components. Additionally, plant safety related functions are not adversely impacted as a result of this activity. Based on the result of the evaluation, implementation of the proposed activity does not involve an Unreviewed Safety Question.

to TXX-98021 Page 11 of 110 Evaluation Number SE 94 095 Revision i Unit: NN2 Activity

Title:

DM 93 062 Installation of Woodward 701 Governor System on Diesel Generators Description of Changes:

Installation of a new control system for the emergency diesel governors such that slow start testing is possible as recommended by NUREG 1431 and GL 84-15. Maintenance and surveillance (most) starts will be done " slow" as opposed to the current " fast" starts. This clow start testing will reduce engine wear and increase reliability. Existing EGA Governors are becoming obsolete and manufacturer support is minimal. The EGA box, MOP and Resister box (existing governor system) will be left in place. A 701 governor, Generator Load Sensor (GLS) and !Aagnetic Pick Up (MPU) Signal Selector (new governor system) will be installed in a new panciln the EDG room. The old components will be left as spares for remaining EGA installations and to avoid requalification of the existing electrical panel. New speed sensors will be added to the EDG. A series of small holes will be added to the shaft in the journal bearing area to accommodate activating the new speed sensors. A new EGB actuator will be installed.

Summary of Evaluation:

A digital control system offers the potential for new failure modes. EMI and common mode failure from software are evaluated as not probable occurrences. The installation addresses possible EMI and the software is in widespread commercial service and has been validated &

verified by the sole source vendor (Cooper) for nuclear service. Impact on the plant is increased reliability of the EDG due to elimination of most fast starts which caused excessive wear on engine components.

Attachment i to TXX-98021 Page 12 of 110 Evaluation Number SE.95-018 Revision 0 Unit: 1N2 Activity

Title:

MM 95-004; DCN 9008 Rev. 0; LDCR SA 95-058; Addition of a Cross-tie Between EFP Discharge Piping & WP Filter Domineralizer Infat Piping Description of Changes:

Plant Minor Modification MM 95-004 added a permanent piping cross tle between the Boron Recycle System (BR) Recycle Evaporator feed pump discharge piping and the Liquid Waste Processing System (WP) Filter Demineralizer inlet piping. This activity allows processing of the Recycle Holdup Tanks (RHTs) liquid waste inventory without the use of temporary drain hoses in the plant. The activity increases personnel safety and improves radiation exposure to meet ALARA guidelines.

The Inventory of the RHTs was originally intended to be processed through the Boron Recycle Evaporator where the boron could be concentrated for reuse. Problems have occurred where lithium also concentrates in the BR Evaporator which in turn can adversely affect the chemistry requirements for the Reactor Coolant. As such, the past handling method for the RHTs was to gravity drain the liquid inventory through the BR Evaporator feed pump casing drain with hoses routed to the Auxillary Building (AB) Sump #7. From AB Sump #7 the inveniory was then pumped to the Floor Drain Tank for processing through the WP Filter Demineralizer as radwaste and discharged. To eliminate the use of hoses described above, this MM provides a cross tie connection between the BR Evaporator feed pumps and the WP Filter Demineralizer feed piping in room X-163, the Steam Generator Blowdown Spent Resin Storage Tank room.

The original option of utilizing the Boron Recycle Evaporator to process the RHT inventory is still available and is not affected by this MM, Summary of Evaluation:

This modification does not adversely affect the performance or function of the BR or WP systems. The double Isolation valves and telltale drain provide positive isolation and leak indication when the line is not in use and also provide a method of draining any dead legs in the piping. The piping associated with this activity is equivalent to the existing piping in room X-163 and is installed in accordance with equivalent requirements. The piping systems are compatible with regard to pressure and temperature rating, piping material specification, Radioactive Waste Management System (RWMS) Jurisdiction and requirements. A review of this activity for potential moderate energy line breaks or flooding scenarios indicated that the System Interaction Program is not affected.

The activity does not involve an Unreviewed Safety Question, require an amendment to the Technical Specifications (TS), increase the probability or consequences of accidents evaluated in the Licensing Basis Documents (LBDs), create the potential for an accident not previously analyzed in the LBDs, or reduce any safety margins existing in the TS bases. This activity does not affect any system used for accident mitigation or affect plarit impact or response to a system failure, and meets all systems design requirements.

- to TXX 98021 Page.13 of 110 Evaluation Number SE-95-033 Revision 0 Unit: NN2 Activity

Title:

OPEN DOOR FOR MAIN STEAM /FEEDWATER VENT PLENUM Description of Changes:

This activity opens the door upstream of the cooling coil associated with the Safeguards MS/FW ventilation plenum in order to provide an alternate path for intake air in lieu of using 100% outside air. Tornado damper CP2-TVSGTD-16 would also be secured in the open position to provide a flow path from the Safeguards 832' elevation. In order to ensure sufficient flow through the tornado damper and into the filtration plenum, one of the two roof ventilators for the Safeguards Electrical Area ventilation system should be turned off.

By opening the door and tornado damper as described above, a MS/FW Area supply air path of less resistance is opened, instead of utilizing 100% outside air, a mixture of outside air and tempered air form the 832' elevation of the Safeguards Building Electrical Area will be used as supply air. This addition of tempered air will allow the air mixture going through the MS/FW -

plenum cooling coil to be cooled further than 100% outside air.

Summary of Evaluation:

The installation of this Temporary Modification does not constitute an Unreviewed Safety Question. The components needed to protect the Train A & B electrical areas from steam and feedwater line cracks or smallline breaks in the penetration room are unaffected by this activity. The Margin of Safety is unchanged. Opening the tornado damper does not increase the heat load handled by the safety related A/C system upon ESF actuation. The open Tornado damper will equalize building pressures as well as a closed tornado damper during a tornado.

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_.- to TXX 98021 Page 14 of 110 -

Evaluation Number SE-95 040 Revision 1 Unit: 1X2

~ Activity

Title:

DMs 95 008,91-137 Rev.1; LDCR SA-96 055,056; IT-96-001,0G2; Reactor Makeup Water 3

System Modifications Description of Changes:

This activity invcives modification of Unit 1. Unit 2, and Common DD Reactor Makeup Water (RMW) system to reestablish expected system parameters (normal operating pressure, flow i

rates) and reduce the maintenance activities associated with the RMW pumps Miniflow lines from the Unit 1. Unit 2, and Common pumps' discharge along with recirculation lines to the RMW Storage Tanks v44 be modified to allow for increased minimum flowrates and more recirculation flow capacity. The safety related recirculation lines will be retained while additional recirculation capacity will take advantage of the non-safety-related cross tie header for the RMW Storage Tanks. Orifices installed for flow restriction at the pumps' discharge will be relocated to the applicable supply lines (s). To reduce the effects of pressure transients i

initiated in another system, a surge suppressor will be installed near the thermal relief valves that are experiencing inadvertent lifting. For the interim between Unit 1 (1RF05) and Unit 2 (2RF03) Implementation, additional operator action will be required to compensate for the unrestricted flow wilenever the Common RMW pump is aligned to supplying Unit 2.

Summary of Evaluation:

In accordance with the FSAR and DBD-ME-028, any credible failure mode of the non-safety, non-seismic cross tie header will be isolatable and will not eliminate the Safety-related protection provided for the Reactor Makeup water pumps via the mlniflow to tank recirculation configuration. All credible failure modes have previously been analyzed and no new failure modes are Introduced. There is no effect on accidents and malfunctions evaluated in the Licensing Basis Documents. There is no potential for creating a new type of unanalyzed event.

There is no impact on ths Technical Specifications or their bases, nor is there any affect on the margin of scfety as a result of this activity or it's implementation, t

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-. -. to TXX-98021 Page 15 of 110 Evaluation Number SE.95-041 Revision i Unit: 1X2 Activity

Title:

Addition of Thermal Dispersion Type Level Switches In ASME Piping Description of Changes:

These design modifications installed level sensing Instrumentation on the steam supply and exhaust lines for the AFWPT 2-01 and on the steam supply and exhaust line of AFWPT 1-01.

The level switches will alarm in the control room if excessive condensation is detected in the lines. The switched initiate no automatic actions except the alarm and are non-1E. The level switches are safety related and seismically qualified for pressure boundary integrity. The alarrn function is required to monitor the standby readiness of the AFWPT. The loss of these alarms during or following any Design Basis Event would not affect AFWPT operability. The steam supply line switches connect to existing instrument taps on the 1/2 MS-27 draln pots. The exhaust side switches required penetrations in the 4"(class 3) exhaust drain line.

The LDCR to the Q list defines the requirements for this new application of this level switch type. Previous types are ASME whereas the new instruments are non-ASME.

Summary of Evaluation:

The safety evaluation concludes that the installation of this levelinstrumentation does riot impact the safety function or operation of the AFWPT or the Main Steam System. The vel switches will enhance the operability and reliability of the system by alerting the operator to a potential problem if excessive condensate accumulates in either of the drain lines. The switches will not affect the normal operation of the AFWPT or Main Steam System because they do not initiate any action other than the alarm and do not affect the margin of safety for these systems.

- to TXX 98021 Page 16 of 110 Evaluation Number SE-95-044 Revision 0 Unit: 1NN Activity

Title:

MM 94 238 R/0(DCN 9572); LDCR SA 95-101; Removal of Steam Generator Channel Head Drain Valve / Tubing & Plugging of Drain Sleeve for Unit 1 Description of Changes:

This activity involves removing valves 1RC-8079A,B,C,D as weil as their associated tubing and fittings and replacing these parts with a plug. These valves were originally installed to allow a simple way to drain any residual water from the steam generator primary channel head prior to opening the primary manway. However this method of draining has not been reliable due to drain line clogging and the residual water in the primary channel head draining through the 3rimary manway. The decision to remove the lines and valves rather than leave them in place s based on two concerns. First of all, these lines have become a crud trap and a significant ALARA challenge when working around them. Secondly, the drain line configuration extends vertically down from the steam generator and is not supported. This configuration can oscillate much like a pendulum during operation and result in accalerated fatigue of the weld which attaches the tubing to the steam generator. For these reasons it was determined to be beneficial to eliminate these lines and plug the penetration at the bottom of the steam generator primary manway, Summary of Evaluation:

The effects this activity has on the plant are not adverse. Among the benefits of this o

. modification are an ALARA savings due to the elimination of a crud trap and the elimination of an identified accelerated fatigue mechanism for the steam generator channel head drain line sleeve penetration. No plant commitmer.ts are affected by this activity. This activity has no impact on the daily plant operations or outage maintenance operations / activities. All original design and safety criteria are met. Based on the evaluat;on, the implementation of the proposed activity does not involve any unreviewed safety questions.

to TXX-98021 Page 17 of 110 Evaluation Number SE-95-056 Revision 1 Unit: 1X2 Activity

Title:

MM 95-216, DCN 9761; LDCR SA-95-122; increase of " Fast" Speed of Fuel Handling Bridge Crane from Controller Programmed Current Max, of 33 FPM Description of Changes:

Increase the " fast" speed of the Fuel Handiliig Bridge Crane from controller programmed current maximum of 33 FPM to new programmed maximum speed of 40 FPM and evaluate any impact this speed increase may have on safety related equipment. (LDCR-SA-95-122; MM 95-216; DCN 9761)

Summary of Evaluation:

The Fuel Handling Bridge Crane is used for handling fuel assemblies within the spent fuel pools, refueling canal, and wet cask pit. Increasing the " fast" speed from t,1e current programmed maximum of 33 FPM to the new programmed maximum of 40 FPM will cause the fuel assembly to travel with more till and will put more force on the top nozzle and facl assembly. Since the trolley hoist can be moved in a direction perpendicular to the bridge movement, using the hand driven chain mechanism, the maximum speed of the fuel assernbly will be slightly higher than 40 FPM. A review was conducted and the conclusion reached that the increased tilt (and resultant swing if the fuel assembly is brought to rest) will not pose any threat to fuelintegrity. Review also concluded that the increased hydraulle loading on the fuel assembly components will be negligible in comparison to fuel assembly handling load allowances. It was concluded that the increased speed will not introduce any significant increase in forces experienced by the top nozzles, fuel assemblies, or the fuel handling tool which could lead to fuel drop accidents. The design basis fuel handling accident as described in FSAR Section 15.7.4 remains bounding regardless of fuel assembly speed. it was also concluded that any moment force experienced by the fuel assembly structure or top nozzle would be insignificant and thus would not be expected to have an adverse effect on the fuel assembly. The bridge stops are modified to stop the bridge crane at 40 FPM as required. it has been verified that any additional deceleration forces which could be experienced by the bridge crane due to the increase in crane speed from 33 to 40 FPM are within acceptable limits.

The maximum stresses in the crane structure during a stop from 40 FPM are within allowable limits and the structuralintegrity of the crane is acceptable. The increase in speed does not represent an unreviewed safety question nor result in a change to the Technical Specifications.

The evaluations and conclusions reached are applicable for t'oth Westinghouse and Siemens supplied fuel assemblies.

_. to TXX-98021 Page 18 of 110 Evaluation Number SE-96 006 Revision 0 Unit: 1N2 Activity

Title:

DM 95 97 " Spare Transformer TXEC1/2" Description of Changes:

This modification provides a bypass power supply for the non-1E load panels XEC1 and XEC2 which will allow maintenance to be performed on the Non 1E transformors TXEC1 and TXEC2.

This bypass power is provided from two transfer switches, (CPX-EPDSNC-03 and 04), fed from the new transformer TXEC1/2 v/hich is fed from non 1E MCC XB2-2. These transfer switches will be normally allgned to their normal source via TXEC1 and TXEC2 and will only be aligned through the new spare transformer TXEC1/2 for maintenance on the normal power source. The transfer to and from transformer TXEC1/2 will be administratively controlled to prevent connection to panels XEC1 and XEC2 simultaneously.

i Summary of Evaluation:

The addition of the non safety spare regulating transformer TXEC1/2 and transfer switches will allow maintenance to be performed on transformers TXEC1 and TXEC2 without interrupting power to the criticalloads fed from priority panels XEC1 and XEC2. This modification will not alter the isolation (dual breakers) from the upstream 1E supply. The existing protection and coordination will be adequate to ensure protectica of the upstream 1E buses from the non-safety supply during transfer. This Maintenance / Operation enhancement will be administratively controlled to ensure transformer TXEC1/2 is properly aligned and controlled to prevent possible extended parallel operation in the event of a failure of a transfer switch. The new equipment will be mounted Seismic Category 11 so as to not impact the Safety systems. All changes in this modification are non safety and are not required to mitigate the consequences of an accident. Based on this discussion and the evaluation herein, no Unreviewed Safety Question exists for this modification.

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- Attachment 1 to TXX-98021 Page 19 of 110 -

Evaluation Number SE-96 020 Revision 0 Unit: NN2 Activity

Title:

Provide tempered air to the MS/FW supply air handling unit.

Description of Changes:

.This activity opens the door upstream of the cooling coil associated with the Safeguards MS/FW ventilation plenum in order to provide an alternate path for intake air in lieu of using 100% outside air. Tornado damper CP2-TVSGTD-16 would also be secured in the open position to provide a flow path from the Safeguards 832' elevation, in order to ensure sufficient flow through the tornado damper and into the filtration plenum, one of the two roof ventilators for the Safeguards Electrical Area ventilation system should be turned off.

By opening the door and tornado damper as described above, a MS/FW area supply air path of less resistance is opened. Instead of utilizing 100% outside air, a mixture of outside air and tempered air from the 832' elevation of the Safeguards Building Electric Area will be used as supply air. This addition of tempered air will allow the air mixture going through the MS/FW plenum cooling coil to be cooled further then 100% outside air.

Tornado damper CP2-0TVSGTD 33 should be opened (at least 4 modules) to ensure that differential pressure due to altered HVAC line up will not preclude the 852' elevation fire door from closing.

Summary of Evaluation:

The installation of this Temporary Modification does not constitute an Unreviewed Safety Question. The components needed to protect the Train A & B electrical areas from steam and feedwater line cracks or small line breaks in the penetration room are unaffected by this activity. The Margin of Safety is unchanged. Opening the tornado damper does not increase the heat load handled by the safety related A/C system upon ESF actuation. The open tornado damper will equalize building pressure during a tornado.

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j to TXX-98021 l

Page 20 of 110 -

R Evaluation Number SE-96-021 Revision 0 Unit: 1NN Activity

Title:

Temp Mod. to provided tempered air to the MS/FW supply air handling unit.

Description of Changes:

This activity opens the door upstream of the cooling coil associated with the Safeguards 1

MS/FW ventilation plenum in order to provide an alternate path for intake air in lieu of using 100% outside air. Tornado damper CP1-TVSGTD-16 would also be secured in the open

- position to provide a flow path from the Safeguards 832' elevation. In order to ensure sufficient flow through the tornado damper and into the filtration plenum, one of the two roof ventilators for the Safeguards Electrical Area ventilation system should be turned off.

By opening the door and tornado damper as described above, a MS/FW Area supply air path of less resistance is opened. Instead of utilizing 100% outside air, a mixture of outside air and tempered air from the 832' elevation of the Safeguards Building Electrical Area will be used as supply air.. This addition of tempered air will allow the air mixture going through the MS/FW plenum cooling coil to be cooled further than 100% outside air.

= Tornado damper CP1-TVSGTD 33 should be opened (at least 4 modules) to ensure that differential pressure due to altered HVAC line-up will not preclude the 852' elevation fire door from closing.

Summary of Evaluation:

The installation of this TM does not constitute an USQ. The components needed to protect the Train A & B electrical areas from steam and feedwater line cracks or small line breaks in the penetration room are unaffected by this activity. The Margin of Safety is unchanged. Opening the tornado damper does not increase the heat load handled by the safety related A/C system upon ESF actuation. The open tornado damper will equalize building pressure during a tornado.

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Attachment i to TXX-98021 Page 21 of 110 Evaluation Number SE-96 024 Revision 0 Unit: 1NN

. Activity

Title:

DM 95 045; installation of Chemical & Volume Cntrl. Sys. Demineralizer Bypass Linellsolation Valve For Cleanup of CVCS Letdown Flow Description of Changes:

This DM installs a bypass line and isolation valve around the CVCS Mixed Bed Demineralizers TBX-CSDMMB-01 and -02. This bypass line will provide a direct full flow pathway to me Boron Thermal Regeneration System (BTRS) Demineralizers TBX TRDMTH 01, -02, -03, -04, and -

05. This modification will provide additional flexibility for cleanup of CVCS letdown flow during periods of Reactor Coolant System (RCS) contamination. Flexibility will exist to allow flow alignment through various combinations of mixed bed, cation bed and BTRS demineralizers.

The BTF 3 demineralizers may be loaded specifically to perform cleanup of CVCS letdown flow.

The CV_S and BTRS functional requirement and capability to provide boration and dilution of the RCS will not be affected by this modification.

Summary of Evaluation:

This modification will provide additional flexibility for cleanup of CVCS letdown flow. The function of the CVCS to maintain the reactor subcritical through boration to compensate for xenon decay and burnout is not affected. Boration in conjunction with the BTRS to accommodate design plant load follow will continue to be available. This modification does not introduced or increase the potential for inadvertent dilution of the RCS. There are no potential failure modes introduced by implementation of the modification. The modification has no effect on accidents and malfunctions of equipment evaluated in Licensing Basis Documents and does not create potential for a new type of unanalyzed event. This activity does not affect any system used for accident mitigation, will not affect plant impact or response to a system failure, and will meet all system design basis requirements. There are no applicable Technical Specifications for the affected components. No plant commitments are affected by this activity. This activity has no adverse impact on the daily plant operations or outage activities.

All original design and safety critoria are met. Therefore this modification does not involve an Unreviewed Safety Question.

Attachment i to TXX-98021 Page 22 of 110 Evaluation Number SE-96-027 Revision 0 Unit: 1NN Activity

Title:

Revision of DBD, Electr. Spec. & FSAR to Add to Appropriate Electrical Equipment Lists Description of Changes:

This activity is for evaluating the impact of not providing separation between associated control cables and non-Class 1E control cables (as well as evaluating adequacy of isolation of cables) at instrument air compressor termination cabinet while implementing the instrument air compressor design modification. (Note that design modification to install larger capacity compressor and the related wiring is evaluated separately under the 10CFR50.59 safety evaluation.

Summary of Evaluation:

- The power for this control circuit originates from the non-Class 1E,750VA,480/120V control power transformer at the new compressor. The transformer secondary is fused with one Bussman FNQ class CC 4A fuse and the primary fused with two Bussman FNQ class CC 4A fuses (one per phase). Since the transformer power is tripped, on any event where Safety injection System (SIS) signalis present, the entire circuit is de-energized and a failure of the non-Class 1E components could not affect the 1E components. Additionally, the worse case -

failure mode for these cables would be a ground fault in the associated cables at the Main Control Board coincident with a catastrophic failure within the non-Class 1E portion of the circuit which creates a direct path to ground from the secondary of the control power transformer. In the unlikely event that this scenario would occur, the secondary 8 A fuse would clear the fault in approximately 0.01 seconds or if the secondary fuse falls to clear the fault, either one of the primary fuses would clear the same fault in 0.025 seconds.

Because these circuits are protected from damaging faults by the inherent characteristics of the circuit design including fuses and the complete circuit is de-energized on a SIS signal, no credible degradation of the 1E system and potential failure modes result by this activity.

This evaluation concludes that the instrument air and the electrical power systems are no:

degraded and the plant is not adversely impacted as a result of the activity. Also the applicable margins of safety are not decreased and the ability of the systems to perform their intended safety function is not compromised by this activity. Based upon the results of this evaluation, implementation of the activity does not involve an unreviewed safety question.

-- to TXX 98021' Page 23 of 110 Evaluation Number SE 96-029 Revision 0 Unit: 1N2 Activity

Title:

Design modification of RCP 1-01/1-04 Smart Motors Replacement Rework; Usage of Galvanized Steel Sheathed Cable as Metal Clad Cable Description of Changes:

RCP motor 1-01 is replaced in 1RFOS with a Smart Motor adding remote visual inspection (RVI) ports and remote monitoring for preventive maintenance intervals. This modification also installs new metal clad cable, cable raceway and connectors in the Unit 1 reactor building for monitoring by pertable test equipn - at provided by Westinghouse during preventive maintenance intervals and troubie shooting activities. Additionally RCP motor 1-04 remote connection box is repositioned to a location where dose rates are minimized.

Summary of Evaluation:

The Smart Motor is the same as RCP motor with additional monitoring capabilities. The Smart Motor monitoring system is a non-safety related passive system that has no impact on the operation of the RCP motor and/or RCS system. This monitoring system will be used to collect data inside containment from the RCP motor sensors using portable test equipment durirg scheduled maintenance intervals and troubleshooting activities. The RCS and RCP motor is described in the FSAR but not to the level of detail for maintenance monitoring. Metal Clad (MC)is described in the FSAR, however this "new at CPSES" galvanized steel sheathed cable is not included in the present description. This addition will require updating the FSAR to allow for the new type of sheathed cable for use now and for future additions. The use of sheathed cable greatly reduces installation time and for this modification reduces doses received inside containment (ALARA). These MC cables carry low level signals, meet IEEE-383 requirements, are installed using Seismic Category ll requirements and meet Regulatory Guide 1.75 separation requirements. The potential failure modes for the Metal Clad cable for faults, open circuits, and cross connection are the same as those enclosed in raceway; therefore installation of these Metal Clad cables do not adversely impact the operation of the units. The additional combustible loading for the exposed conductor for air drops is negligible and has been evaluated as not an impact to the plant fire safe shutdown capability.

Based upon the results of this evaluation, implementation of the proposed activities dot not involve any unreviewed safety question.

_ to TXX-98021 Page 24 of 110 Evaluation Number SE-96-032 Revision 0 Unit: 1X2 Activity

Title:

Physical Separation of Filtered Water Strg. Tank from filtered Water Clearwell and Abandonment / Removal of Turb. Bldg. Water Trtmnt.Eqpmnt.

Description of Changes:

.The Filtered Water Clearwellin the 778 level of the Turbine Building has been "abandoried in place", The line between the Filtered Water Clearwell and the Filtered Water Storage Tank will be cut and capped under the Maintenance Activity Program to provide a physical separation between operating and out-of service Water Treatment equipment.

The Fittered Water Clearwell, N.,assified Water Clearwell, and related Reverse Osmosis, Demineralizer and Chemical injection equipment on the 778' level of the Unit 1 and Unit 2 Turbine Building is abandoned in place and will be physically removed at a later date.

Summary of Evaluation:

The abandonment of the original water treatment equipment and its physical separation from the Filtered Water Storage Tank has no impact on the design or operating parameters of any.

equipment important to safety. Interactions with operating systems and the isolation details for each particular system are discussed.

This evaluation has determined that this activity does not present an unreviewed safety

question, no new accident or malfunction of equipment other than previously evaluated has

' been introduced, and the margin of safety as defined in the Technical Specifications has not been reduced.

_ _.__ to TXX 98021 Page 25 of 110 Evaluation Number SE-96 034 Revision i Unit: 1NN' Activity

Title:

Design Modification For Replacement of 10kva & 7.5 kva inverters and 225 amp Chargers with SCI Model 10 Kva Inverters and 300 amp Chargers Description of Changes:

The design modification involves replacement of Class 1E 10 kva Elgar and 7.5 kva NSSS Westinghouse inverters with Class 1E 10 kva SCIinverters and the replacement of their respective Class 1E 225 amp battery chargers with Class 1E 300 amp SCI chargers; It also involves addition of Class 1E 10 kva SCIinverters as installed spare per safety train and other associated configuration changes including changes to Inverter / battery charger circuit breaker ratings. The modification additionally includes replacement of Non-Class 1E 10 kva Elgar inverters with 10 kva SCI inverters. The modification also deletes 480 V ac feed to the Class 1E and non Class 1E inverters, adds hydraulle lift platform in Unit 1 cable spread room for equipment handling and relocates UPS fan coil units within their respective rooms.

4 Summary of Evaluation:

The design modification activities have been reviewed for impact on fire safe shutdown analysis; fire protection program; seismic interaction; system interaction; plant safe shutdown; emergency diesel generator loading; electrical protection, separation & isolation; acceptability of electrical equipment ratings and for voltage adequacy for equipment operation. Based on this review, no credible failure modes associated with the design modification activities were identified.

The impact of greater heat losses with the new equipment was reviewed and determined that the slight increase in neat loads is within the capability of the UPS HVAC system. Therefore, the qualified life and operation of the equipment in the UPS rooms are not affected during all modes of plant operation (normal operation, Loss Of Offsite Power and Station Blackout conditions).

The design modification implementation plan described compensatory actions requiring the temporary sealing of any unattended breached EQ barriers. Upon review of this plan, it was determined that these actions are acceptable and satisfy the CPSES EQ Barrier Program.

Implementation of this modification requires CPSES Technical Specification changes that include changes to the battery charger surveillance parameters, credit for using spare inverter and inverter nomenclature changes. These changes are addressed in the Technical Specification Change Request TS96-005 and the License Amendment Request LAR 96-005 which has been approved by the NRC.

Based on the 9bove evaluation,it is concluded that the implementation of the DM does not result in an unreviewed safety question.

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.... - _ - to TXX-98021 Page 26 of 110 Evaluation Nu'nber SE-96-038 Revision 0 Unit: 1NN Activity

Title:

Design Basis Change to Allow Installation of an ASME Code Class Blind Flange at the Outside Containment Isolatinn Valves Description of Changes:

The new design provides for the installation of an ASME code class blind flange outside contalnment to seal off the penetration that has failed to pass the LLRT test. The design basis provides the justification for using the blind flange as a repair to the outside containment isolation valve and as a safety measure if the inside containment valve falls the LLRT test. The blind flange will be added to the FSAR.

Summary _of Evaluation:

The evaluation determined that the blind flange is an acceptable substitute for the outside containment valve under the Technical Specifications. The seal provided by the double "O" ring on the blind flange is equal to or better than the seal provided by the rubber seat mLierial in the containment isolation valve.

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- Attachment 1 to TXX-98021 Page 27 of 110

- Evaluation Number SE-90-039 Revision 0

- Unit: NN2 Activity

Title:

TM 2 96 008 Revision 0;Pressr. Relief Valve Addition to Maintain Si Hdr.Pressr.Below Existing Relief Valve Set Point / Monitor Leak Off Rate Description of Changes:

This Temporary Modification documents the use of temporary relief valve, drain valve, flow monitoring devices and associated piping, tubing, and supports on the Safety injection (SI) header. The installation will compensate for increased pressure in the Si caused by RCS leakage into the system and will be used to provide a pressure relief mechanism. This will allow the SI system pressure to be controlled to avoid challenging the existing relief valves to-operate at the low mass input of the RCS system and improve the ability of the relief valves to reseat at pressures below the set point. The leakoff rate from the relief valve will be monitored to prevent draining the Sl outside of containment at a volume or rMe greater than allowed by the Tech Spec or FSAR. The spare relief valve that will be used for this TM will be modified by Installing new spring and washer internals, rerating the valve to Si system requirements, and establishing a new relief set point pressure. A new nameplate is required for the rerated relief valve in accordance with site procedures.

Summary of Evaluation:

2 The installation of this Temporary Modification will provide the ability to limit the SI system header pressure and monitor system leakage. The piping installation has been analyzed for seismic and stress considerations and meets applicable design requirements for permanent plant installations. Consideration has been given for all potential failure modes and it has been determined that failure modes that affect safety significant components can not increase the likelihood, severity, or consequences of any accident analyzed in LBD's. This Safety Evaluation has determined that the installation of the relief valve, drain valve, piping, tubing, and supports does not involve an Unreviewed Safety Question or require an amendment to the Technical Specifications. This activity does not adversely affect any system used for accident mitigation, will not impact plant response to a system failure, and will meet all system design basis requirements.

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. Attachment 1 to TXX 98021 Page 28 of 110 Evaluation Number SE.96 040 Revision 0 Unit: 1NN Activity

Title:

Replacement of Gland Steam Condenser Bypass Flow Control Valve 1-FV-2243 with An Orifice Plate Description of Changes:

This activity included removal of the Gland Steam Condenser Bypass Valve 1-FV-2243 and associated controls and alaims and Installing a flow restricting orifice on Unit 1. Manual gate valve 100 0255 was replaced with a V-Ball valve, and a bypass valve was Installed across gate valve 1CO 0256. In addition, manual valve 1CP-0001 on the discharge side of the Backwash Recovery Tank A was replaced with an air operated valve. A new non-safety related motor control center (MCC XB1-8)is provided to feed the new loads for this upgrade.

Summary of Evaluation:-

The replacement of valve 1-FV-2243 with a restricting orifice plate is acceptable. The evaluation determines that the orifice plate will distribute required design flow to the associated components without adversely affecting any of the associated systems or components. This activity does not introduce any new failure modes for the plant or any associated systems. The orifice plate provides the same function as the installed valve, and as a passive component it is not credible that the orifice will fall and adversely affect any system, structures or components.

No credible failure modes are introduced by this modification. The new MCC will provide power for the new motor operators and programmable logic controller for Unit 1 Condensate Polishing System. The addition of the new electricalloads has been evaluated per Safety Evaluation 94-007.

k to TXX 98021 Page 29 of 110-Evaluation Number SE-96-041 Revision 0 Unit: 1N2 Activity

Title:

Unit 1 TDAFW Pump Steam Supply and Drains Modification Description of Changes:

This modification to the steam supply lines and associated drains to the Turbine Driven Auxiliary Feedwater Pump turbine is intended to reduce overspeed trips and increase the reliability of the TDAFW pump. The " tee" connection and upstream check and block valves for the two steam supply lines to the TDAFW pump are being relocated from the pump room to the main steam and feedwater penetration area to reduce the volume of pipe between the steam 1

admission valves and the turbine. One of the existing supply lines will be retained as the common header and will have knockout pt,ts installed to remove condensation generated by cold start transients. The other supply line will bc converted to a drain / vent line for the knockout pots and will be routed to a new flash tank. The existing high and low pressure drains will also be routed to the flash tank which will be vented to atmosphere and drain to the floor 2

- drain system. Additionally, the governor valve stem drain will be rotated to improve drainage.

Level instrumentation is being added to the turbine discharge drain.

Summary of Evaluation:

This modification will reduce the volume of condensation gen 3 rated and will remove the remaining water before it passes through the turbine resulting in smoother starting transients and a lesser demand on the governor to prevent over speed trips. The flash tank will ensure that the steam from the drains and knockout pots do not create a harsh environment for the Class 1E equipment in the area. The mods will be in accordance with the applicable codes and standards per the FSAR. The overall effect will be improved reliability of the TDAFW pump.

There is no unreviewed safety question concerning these mods.

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. to TXX-98021 Page 30 of 110 Evaluation Number SE 96-042 Revision 0 4

Unit: 1NN Activity

Title:

Temp f3od to Support 1RF05 Installation of Unit 1 Design Mod for RPS Testing in Bypass, including the RCP Set Racks & N 16 Cabinets Description of Changes:

This temporary modification is to support the installation of the Unit 1 Reactor Protection System Testing in Bypass design modification during the 1RF05 outage. This TM involves the monitoring of certain instrumentation needed to perform surveillences required per Technical Specifications during' Modes 5 & 6 for Unit 1 and Mode 1 for Unit 2. In addition, the Reactor Coolant System loop 1-01 wide range pressure and the Steam Generator wide range level monitoring, which are not required per Technical Specifications while in Modes 5 & 6, have been included per Operations request. The implementation of this TM will provide continuous display of the process parameters on the main control board in the control room and on the plant computer as applicable. This TM must be installed and be functional prior to the de-energization or the reactor protection set racks TBX-XIELRK-01 -02,-03,-04, and N-16 cabinets TBX-XIELSS-501X & TBX-XIELSS 50X.

Summary of Evaluation:

The evaluation concludes that implementation of the proposed temporary modification is acceptable because it provides the capability to comply with Technl. Specifications during the 1RF^5 outage of Unit 1 and continued operation of Unit 2 in Mode 1. This activity is acceptab c because it temporarily provides the capability of monitoring essential process paramete,: luring the 1RF05 outage to support the installation of Unit 1 Bypass Modification DM 93-067. This activity presents no new failures modes for the plant or any plant systems.

The implementation of this activity will facilitate the monitoring of system status by the control room operators during Modes 5 & 3 because it will provide continuous display of the required channels in the control room. No credible failure modes are associated with this temporary modification.

-. to TXX-98021 Page 31 of 110 Evaluation Number SE-96-044 Revision 0 Unit: 1NN Activity

Title:

Design Modification for installation of 480 V_ Feedor to Power Personnel Airlock from MCC 1EB1-2 Tripped by SI Signals Description of Changes:

A 480V feeder is provided to the personnel airlock from MCC 1EB1-2. Power at this MCC is de energized upon receipt of a Train A 51 signal. A 480V shunt-trip breaker, which de-energizes the personnel airlock power feeder upon receipt of a Train B Si signal, is instelled in a panelin the power feeder between MCC 1EB1-2 and the personnel airlock. An administratively-controlled 480V manual transfer switch is installed in the power feeder between the MCC 1EB1-2 and the shunt-trip breaker which provides an alternate power source from the onsite 25kV loop during a Train A bus outage. The 25kV loop feed is protected by two breakers in series and included in the TRM to ensure surveillance requirements of Technical Specifications 4.8.4 (penetration protection) are met. Also Licensing Basia Documents are updated to reflect the design and to justify separation requirements.

Summary of Evaluation:

The safeny-related purpose of this design modification is to remove power to the airlock automatically on the receipt of an SI signal. Separation requirements within the shunt trip breaker panel are justified by failure analysis. Bypass of trip signals by transfer switch is procluded by restricting its use for Modes 5 and 6. Failure modes identified are analyzed and found to be either precluded by design or with alternative action available. Also basis for Technical Specifications are not impacted.

Based on the results of the safety evaluation, implementation of the activity do not involve an unreviewed safety question.

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_ to TXX-98021 Page 32 of 110 Evaluation Number SE-96-045 Revision 0 Unit: 1NN Activity

Title:

Replacement of Unit 1 Feedwater Pump Turbine Analog Control System with a Digital Control System (GE MARK V)

Description of Changes:

This modification replaces the existing analog electronic governor control for each turbine driven feedwater pump with a shared digital system (GE Mark V). The replacement is an acceptable substitute, however, it is not functionally or physically interchangeable with the original analog system and is a modification to the plant. This modification consists of adding an air conditioned room, removing the existing electronic governor from the SG FW Pump Turb.

(SGFPT) skid and replacing it for each pump with a shared digital governor GE Mark V system to be located in the room. Additional sensors are added to provlds a 2/3 logic for the new digital governor. SGFPT sensors will have time delays provided by the Mark V system. This modification to the SGFPT is desirable for trip reduction. SGFPT sensors will be routed through the Mark V with the exception of the trip solenold. The trip solenoid and overspeed

' features are unaffected. The Mark V will have a different power supply than the original electronic governors. The existing manual control in the main control room will t,e replaced with a five position switch. The operator will now be able to respond to trouble alarms using one of the two interface computers, one of which will be located in the new room in the turbine building and the other in the Unit 1 main control room computer room. The design addressed concerns specific to the new digital electronics that could result in failure modes (use of common software in redundant channels and increased sensitivity to the effect of electromagnetic interference) and system malfunctions that were not considered in the axisting design.

Summary of Evaluation:

This modification is being made to reduce the probability of a malfunction of the feedwater pumps. Therefore the probability of a malfunction or an accident is decreased. This modification does not create the possibility of a new failure mode since the accident analysis bounds the failure modes of the new design. The accident analyses remain bounding for this i

change to the facility. There are no technical specifications affected and there is no reduction in the margin of safety. Therefore, the modification does not constitute an unreviewed safety question.

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l to TXX-98021 Page 33 of 110 Evaluation Number SE-96-047 Revision 0 Unit: 1N2-Activity

Title:

TM 1-96 011; LDCR SA-96-088; Use of Temporary Conta!nment Equipment Hatch Cover Description of Changes:

An outage equipment hatch with pipe penetrations is to be installed to allow steam generator activities concurrent with fual mwement. The outage equipment hatch is in two sections. One section is for the penetrations and is installed when containment closure is not required. When containment closure is required, the second piece of the outage equipment hatch is installed and the entire equipment httch opening is covered by the two pieces. All penetrations and joints associated with the outage equipment hatch shall be sealed. The equipment which passes through the outage equipment hatch penetrations will not provide a path between the containment atmosphere and the outside atmosphere. With the exception of loss of shumown cooling where containment pressures and radiation levels are potentially much more severe than a fuel handling accident, the outage equipment hatch provides a fission product barrier which is equivalent to the permanent equipment hatch. Since the outage equipment hatch can not be relied upon for loss of shutdown cooling all penetrations are equipped with quick disconnects to allow timely closure of the equip 9 ant hatch. This evaluation is valid for all activities during Modes 5,6 and refueling. The TM is used for the physical details, the LDCR updates the FSAR to describe this activity.

Summary of Evaluation:

There are no credible failures introduced by this temporary modification to the equipment hatch penetration as described in the FSAR. Technical Specification 3.9.4 is to be satisfied while this modification is in place. There is no reduction in the margin of safety because the containment closure function is te be maintained during core alterations and movement of irradiated fuel during Mode 6.

I Attachment i to TXX 98021 i

Page 34 of 110 i

Evaluation Number SE. 96 048 Revision 0 1

Unit: 1NN Activity

Title:

Moddication of Containment Penetration for Tubing and Cables for Unit 1 Refueling Outage i

1RFOS Description vf Changes:

Blind flanges from two of three spare pipes in a con'ainment mechanical penetration were removed and LLRT tubing and telecom cables were cassed through while in MODE 5. Prior to MODE 6 app!Icability of Tech Spec 3.9.4. the pipes were sealed with a suitable penetration seal which ensured the containment ctmosphere was isolated from the outside atmosphere. The tubing was isolated in accordance with TS 3.9.4 during core alteration and fuel movement l

Inside containment.

Summary of Evaluation:

There are no credible f ailures introduced by this modification to the containment penetration as kacribed in the FSAR. Technical Specification 3.9.4 was to be satisfied while this modification was in place. There was no reduction in the margin of safety because the containment closure function was maintained during core alterations and rnovement of irradiated fuel during MODE 6.

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I Attachment i to TXX 98021 Page 35 of 110 l

Evaluation Number SE. 96 049 i

Revision 0 Unit: NN2 Activity Title ONE 96 726; Operation of Safety injection Pump Discharge Piping Above Design Pressure of 1

1750 psig l

Description of Changes:

The design pressure of the Si pum) discharge piping is 1750 psig. RCS leakage into the system through the pressure isolat on check valves is pressurizing the discharge piping and components above the design pressure. The discharge header is expected to rema n 1

pressurized until check valve repairs are accomplished no later than 2RF03.

Summary of Evaluation:

The piping installation has been analyzed for seismic and stress considerations and meets app!! cable design requirements. The design pressure in this portion of the SI system is 1750 pulg With the 10% accumulation value of the relief valves, the piping could be exposed to pressure conditions up to 1925 psig. Analyses have been performed for this portion of the system to 2000 psig Consid6 ration has been given for all potential failure modes and it has been distermined that there are no credible failure modes that could affect components-1 i

resulting 1 an increase in the probability, severity, or consequences of any ticcident analyzed In LBDs. 7he activity does not involve an Unreviewed Safety Question or require an 1

amendment to the Technical Specifications. This activity does not adversely affect any system used for accident mitigation and will not impact plant response to a system failure.

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.. _ to TXX 98021 Page 36 of 110 Evaluation Number SE- 06 050

{

Revision 0 Unit: 1N2 Activity

Title:

LDCR IT 96-003; Revision to inservice Test. Plan for Addition of Relief Valves Protecting Systems / Components Performing Safety Function Description of Changes:

The activity being evaluated is a change to the CPSES Unit 1 and Unit 2 Inservice Test Plan (IST). This change is the addition of existing ASME Code Class 2 and 3 relief valves that protect systems and components which:

1) Perform a specific function in shutting down a reactor to the safe shutdown condition, or i

2) maintain the safe shutdown condition, or

3) mitigate the consequences of an accident Summary of Evaluation:

This change ensures readiness of relief valves that protect ASME Code Class 2 & 3 systems and components. As such this change will not create a new accident, nor Inc ease the probability of an existing licensing basis accident, nor decrease the existing margin of safety.

This change will ensure compliance with the current NRC/ASME philosophy of the scope of relief valves in the IST Plan.

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Attachment i to TXX 98021 Page 37 of 110 Evaluation Number SE.06-053 Revision 0 Unit: 1NN i

Activity

Title:

Provide temp. power to plant eqpmt during the implem of DM 96 0013. SCl Inverter Acceptance Tests Description of Changes:

During the implementation of DM 96 0013," Replacement of the Class 1E inverters and Battery Chargers," temporary power is required to support the design modification's implementation schedule. TM 106 0015 provides temporary power to DC panels 1ED1-1 and 1ED12 utilizing the prototype SCI charger (procured for equipment qualification testing) and non class 1E battery charger BC1D24. TM 196 0016 temporarily powers DC panels 1ED21 and 1ED2 2 using the same chargers. TM 196 0018 provides temporary 480V power from the 25kV plant i

support system to power two temporary chargers that will be used to perform maintenance testing on the class 1E batteries during the new battery chargers' acce stance testing. Test procedures PPT TP 96A 11,12 and 13 temporarily uses non class 1E Dattery charger BC1D24 3

for SCI inverter acceptance testing.

Summary of Evaluation:

The temporary modifications and test procedures are allowable and acceptable per plant design and licensing basis documents. The temporary modifications are required to implement DM 96 0013. The test 3rocedures are required to perform acceptance testing for newly installed Inverters by D W 96 0013.

The activity is consistent with the existing licensing basis and, as such, does not aff6ct the 4 -

probability or consequences of accidents or malfunctions of equipment important to safety analyzed in the LBDs, does not create the possibility analyzed, and does not require a change to the TSs. The activity makes power supply changes (temporarily changing facility drawings) for the duration of each installed temporary modification or test procedure performance.

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- to TXX 98021 Page 38 of 110 f

Evaluation Number SE 96 056 Revision i Unit: 1NN Activity Title.

CPSES 1. Cycle 6 Coro Configuration j

4 Description of Changes' During the refueling outage for CPSES 1 (1RF05), prior to operation of Cycle 6, eighty four fresh Region 8 fuel assemblies manufactured by SPC will replace eight Region 4B assemblies and seventy six Region 6 assemblies. Also, one Standard (STD) fuel assembly manufactured by Westinghouse Region 3 assembly will be replaced with another Region 3 assembly from Cycle 2.

Summary of Evaluation:

This mixed core configuration has been evaluated for mechanical and thermal hydraulle compatibility between the SPC and Westinghouse fuel assemblies. All applicable oesign 1

criterla were determined to be satisfied. The neutronic characteristics of the Cycle 6 core -

configuration have been evaluated for their effect on the accident analyses, in all cases, it was determined that the applicable event acceptance criteria are satisfied. Because all mechanical design criteria continue to be satisfied, there is no reduction in any failure point introduced by the Cycle 6 core configuration. All acceptance criteria of the accident analyses continue to be satisfied; therefore, there is no increase in the consequences of any accident previously analyzed. Based on the foregoing, it is concluded that the Cycle 6 core configuration does not reduce any margin of safety as defined by the plant Technical Specifications; therefore, the proposed change does not involve an unreviewed safety question.

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Attachment i to TXX 98021 -

Page 39 of 110 Evaluation Number SE 96-057 Revision 0 Unit: 1X2 Activity

Title:

Revision of Category i Backfill minimum depth requirements as shown in FSAR, Figure 2.5.4-32.

Description of Changes:

Revise the minimum depth requirement of Category I backfill over safety related commodities from 4'6" to 2'6" as shown in FSAR, Figure 2.3.4 32.

Summary of Evaluation:

The described activity will be performed based on a calculation which provides the engineering basis for Tornado Missile protection of underground safety related commodities. The activity does not involve an Unreviewed Safety Question, change any regulatory commitments, or require an amendment to the Technical Specifications.- This activity does not adversely affect any system used for accident mitigation nor does it impact plant respanse to a system failure.

This change meets all design basis requirements.

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t to TXX 98021 Page 40 of 110 i

1 Evaluation Number SE 96 058 Revision 0 Unit: 1N2 Activity

Title:

Procedure No. PPT-S12701; Safety injection Accumulator Upstream Check Valve Open Test (8956A, B, C, & D)

Description of Changes:

The subject valves are classified as active valves and must be tested for operability per the inservice Testing Plan. In lieu of full stroke exercise testing, an alternate method of i

disassembly and inspection is currently used. This safety evaluation supports the performance of flow testir:g and nonintrusive techniques. A procedure has been prepared to perform a full stroke open test, utilizing the accumulator tank discharge line to open the valves. Acoustic emission techniques will be used to monitor valve disc position during the test.

Summary of Evaluation:

Compliance with ASME Code Section XI and 10CFR50.55a requirements, as defined by CPSES procedures, was venfled. The effects of the check valve test on affected components such as the accumulator tanks, piping, reactor vessel, and water chemistry has been evaluated, impact on the Safety injection system or other systems whose operation is required and may be affected by the performance of the test, have also been evaluated. System availability is not affected since the test will be performed either in Mode 6 or with the core offloaded. The accumulators are required to be operable for injection only in Modes 1,2, and

3. There are no hardware changes and the test parameters are within the normal operating parameters of the accumulator injection system.

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__- _._ -. _ - _ _.. _.. to TXXo98021 Page 41 of 110 Evaluation Number SE 96 059 Revision 0 l

Unit: NN2 Activity

Title:

i DCN 10507 Revision 0; LDCR SA 96 099; Bypassing of Reactor Coolant Pump Motor 2 04 Control Room i

Ammeter Due to Faulted Cable Description of Changes:

The DCN 10507 provides details and justification to lift cable NK203589 from terminal block l

AK 9 and AK-10 in accordance with the specification ES 100 and to install a jumper across these terminal points. This activity is being performed to prevent equipment damage due to faulted cable between the non safety switchgear 2A4 anc the control board 2 CB 05.

Summary of Evaluation:

The faulted cable NK203589 provides signals for a remote ammeter 2 ll RCP4 in the control room and to a transducer A XD 2/2PCPX4 providing inputs to the plant computer. With the determinated and the current transformer burden is reduced. All components af(fected a of the reactor coolant system (RCS) and the affected equipment are non safety related. Th s activity prevents remote current indicator readings for the affected RCP motor both on the control board and the plant computer. However, alternate operator Indications are available to include control room RCP breaker 2 04 position Indication and RCP flow Indication; RCP motor t

ctrrent readings are still available at switchgear bus 2A4.

RCPs are required to c aerate only under normal plant conditions and are not required to operate during an accicent to meet safety functions. The loss of Indication to an RCP motor does not affect probability or c sequences of accidents or malfunctions of equipment important safety analyzed in f

.icensing basis documents, does not create the possibility of accidents or malfunctions of equipment important to safety diff3 rent from previously analyzed.

Based upon the results of this evaluation, implementation of the proposed activity does not involve an unreviewed safety question.

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_ _ _. to TXXo08021 Page 42 of 110 Evaluation Number SE 96 060 Revision 0

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Unit: NN2 l

Activity

Title:

TM 2 96 008 Rev. 2; Addition of Pres. Control Device / System to Maint. SI Headr. Pres. Below Norm. Relief Valve Set Point / Monitor Leakoff Rate Description of Changes:

This Tem)orary Modification documents the use of a pressure control system which utilizes flow restr cling tubing, drain valve, flow monitoring devices, connecting piping and tubing, and associated supports on the Si header and returns the leakoff flow to the Boron recycle system.

The installation will compensate for the increased pressure in the Si header caused by RCS back leakage into the system and will be used to 3rovide a pressure relief mechanism. This will al!ow the Si system header pressure to be conito led to avoid challenging the existing relief valves to operate at the low mass input from the RCS and improve the reliability of the relief valves by reducing system pressure below their set point. The leakoff rate will be monitored to prevent draining the SI system outside of containment at a volume or rate greater than allowed by Technical Specifications of the FSAR.

Summary of Evaluation:

This installation of this Temporary Modification will provide the ability to control and thereby limit the SI system header pressure and monitor system leakage. The piping and tubing installation has been analyzed for seismic and stress considerations and meets app!! cable design requirements for permanent plant installations. The flow restricting tubing has been sized and analyzed to limit leakoff flow to well within the margins assumed in the LDB's.

Consideration has been given for all potential failure modes and it has been determined that faliure modes that could affect safety significant components can not increase the like!ihood, severity, or consequences of any accident analyzed in the LDB's. This Safety Evaluation has determined that the installation of the piping, flow restricting tubing, drain valve, piping, supports, connecting tubing, and flow metering devices does not involve an Unreviewed Safety Question or require an amendment to the Technical Specifications. This activity does not adversely affect any system used for accident mitigation, will not impact plant response to a system failure, and will meet all system design basis requirements.

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XX 98021 Page 43 of 110 Evaluation Number SE. 96 061 Revision 0 Unit: 1X2 J

Activity

Title:

Deletion of the Requirement for Performing Resistance to Ground Checks on Signal Cables During Regular / Scheduled Outages

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Description of Changes:

This LDCR is generated to delete the third paragraph on page 7.3 77 of the FSAR. The existing paragraph Indicates that maintenance checks for resistance to ground are performed on the signal cables during regularly scheduled refueling outages because of degradation due l

to thermal and radiation environments. This type of practice is not necessary because of the Environmental Equipment Qualification program mandated by 10CFR50.49 The EQ program t

at CPSES identifies, based on type testing,if any maintenance checks are required to support I

a 40 year plant life.

Summary of Evaluation:

Allinstrumentation cables, at the time of being envir9nmentally quallfled, were subjected to 40 years of accelerated therrnal and radiation aging. in addition, these cables were also subjected to the worst of LOCA/MSLB environments (temperature, pressure, humidity, chemical spray, etc.). During the LOCA/MSLB test, resistance to ground tests are performed. The worst case demonstrated values of resistance to ground are then compared to acceptance criteria of the test and factored into the loop accuracy of the circuit. This means that the resistance to ground values will not degrade before 40 years of operation. Environmental qualification tests (which Identify what constitutes acceptab e thermal, radiation, etc... degradation) are incorporated into EQ summary packages. Maintenance activity based on these EEQSP's are then identified via the EQ maintenance manual (EQMM) for incorporation into the CPSES maintenance program.

Therefore, it is not necessary to perform a resistance to ground check because it has been demonstrated via type test that the values will not be lower than tested values. Additionally, 10CRF50.49 requires that maintenance / replacement be performed on the expired quallfled life but it does not require any surveillance check to be performed.

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to TXX 98021 Page 44 of 110 Evaluation Number SE 96 062 Revision 0 Unit: 1NN Activity

Title:

TM 19ti 0019 R/0; Temporary Power for Battery Charger BC1D2 During the B Train Bus Outage in 1RFOS Description of Changes:

The temporary modificattun is for providing tempolary power to battery charger BC1D2 during the B train bus outage in 1RFOS in order to maintain power to the 125/250 Vdc loads and to maintain the battery in a fully charged condition. This temporary modification applies only

- during the B train bus outage when the plant is in MODE 6.

Summary of Evaluation:

The use of existing plant support power feed to battery charger BC1D4 to temporarily supply BC102 in parallel when the battery is fully charged and the bus loads are minimal, is acceptable for the duration of the B train bus outage during MODE 6 in 1RFOS. The FSAR discussion, figures and DBD-EE-44 and 57 are impacted temporarily by this modification but does not create any new failure modes. Regulatory Guide 1.75 separation requirements are maintained. As long as the Regulatory Gulc.e 1.75 separation is maintained, there is no potential safety impact since the non safety battery chargers are not required for safe shutdown. System Interaction, fire safe shutdown and commodity clearance issues are not applicable during MODE 6. The plant support power feed has been previously evaluated as acceptable for use during outages. The potential failure modes for the cable and connections is the same as the existing installation (i.e faults, open circuit, and cross connection) which has been previously analyzed. Therefore this temporary modification does not involve an unreviewed safety question.

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Attachment i to TXX 98021 l

i Page 45 of 110

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i Evaluation Number SE 96 063 Revision 0 f

Unit: 1N2 Activity

Title:

Incorporation of Electrical Reliability Council of Texas OperatinD Guide Changes to Reflect independent System Operator Description of Changes:

l FSAR section 8.2 is revised in accordance with the Electrical Reliability Council of Texas i

(ERCOT) Operating Guides for responsive reserve obligations, and to provide clarification for maximum and minimum grid voltages at CPSES.

Summary of Evaluation:

The changes in ERCOT operating guides to incorporate Independent System Operator (ISO) functions would not affect grid stability and reliability. The ERCOT operating reserve requirements are more restrictive than the concepts in the NERC operating manual. ERCOT responsive reserve requirements are based on system reliability studies. The minimum responsive reserve capability maintained by ERCOT is sufficient for security of the system to maintain a high system reliability. The reliability and availability of offsite power source for safe 4

shutdown of CPSES will not be adversely affected by the revised ERCOT guides.

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Attachment i to TXX 98021 1

Page 48 of 110 Evaluation Number SE 96 064 i

Revision i Unit: 1X2 Activity

Title:

Revise Large Break LOCA Analysis with Temporary Assessment for Unit 1 Cycle 6 and Unit 2 Cycle 3.

Description of Changes:

The Unit 2 Cycle 3 large break LOCA was re analyzed using the previously approved computer t

code (TOOD IE-2), per a commitment to the NRC. Bnth the Unit 1 Cycle 6 and Unit 2 Cycle 3 large break LOCA analyses were revised to include temporary assessments of the effects of a t

"non physical" behavior in the TOODEE 2 computer model. Also, the Unit 1 Cycle 6 and Unit 2 Cycle 3 large break LOCA analyses were revised to include an error correction to one of the automated codes used to complete to large break LOCA evaluations.

j Summary of Evaluation:

The Unit 2 Cycle 3 large break LOCA was re analyzed using the previously approved computer i

code (TOODEE 2) and all event acceptance criteria weie found to be satisfied. Also, s tem)orary assessment has been assigned against both Unit 1 Cycle 6 and Unit 2 Cycle 3 ana yses, pending resolution of the non physical behavior in the TOODEE 2 computer code.

No Fq limit reduction was needed and all event acceptance criteria are satisfied.

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.... to TXX 98021 Page 47 of 110 Evaluation Number SE 96 065 Revision 0 I

Unit: 1NN Activity Title' Effects of FW Venturl Fouling on Unit 1 Cycle 6 Uncertainty Analysis Description of Changes' The activity permits operation of CPSES 1 during Cycle 6 with a detectable amount of venturl j

fouling. The primary, non conservative effect of the fouling is to increase the uncertainty associated with the measurement of the precision RCS flow measurement.

4 Summary of Evaluation:

The RCS flow measurement uncertainty, considering the effects of the feedwater venturi l

fouling, is calculated to be 1.95% flow. This value is greater that the 1.8 flow uncertainty -

currently provided in the required RCS flows provided in Technical Specification 3.2.5. The additional uncertainty (above the 1.8% flow) must be accounted for prior to comparison with the Technical Specification required flow,in a manner similar to what is done for the effects of the lower plenum flow anomaly. Compensatory actions were proceduralized in the appropriate test prot.edure.

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_ -.__ _ _._ _._._ _ _ _ _ _ ___..___._ _ _. to TXX 98021 Page 48 of 110 Evaluation Number SE-96 066 Revision 1 Unit: 1N2 Activity

Title:

DM 95 081. Providing Spent Fuel Pool (SFP) Cooling Pump Protect, on Low Low SFP Level; j

Providing "2 out of 2" Logic for Pump ProtJLevel Indic.

Description of Changes:

The ex sting level instrumentation for the Spent Fuel Pools piovide a high and low poollevel I

alarm along with a LO LO Spent Fuel pool cooling pump trip. These instruments will be replaced with level instruments that will retain the function of high and low level alarms and add i

the function of providing pump protection for low low pool level with 2 out of 2 logic. The 2 out of 2 logic will ensure that a single instrument malfunction will not be the cause oL a pump trip.

In add tion this activity adds levelindication of the spent fuel pool to the plant computer. This

. will allow the control room to monitor level of the fuel pool. A non conforming condition on the existing levelInstrument stilling well will be resolved as part of this activity. The new stilling well will be properly classified and the suppcrt will conform to site standards.

i Summary of Evaluation:

The new Instrumentation will be similar to the existing devicer The existing instrumentation will be slightly modified to provide the continuous output to the plant computer and the 2/2 trip logic. The sensors will be installed in the same location as the existing serisors and a duplicate electronics box which contains the terminals and relays for the switch functions will be installed in the S pent Fuel Pool Pump rooms. Class 1E and Non Class 1E cables terminate on the same terminal block inside the electronics boxes. This condition was previously evaluated and documented in the FSAR. A quallfled isolation device is used to isolate a Class 1E signal to a Non Class signal to the plant computer. The new instrumentation configuration is seismically quellfled. These changes will not degrade the operation of the Class 1E boxes. The alarm set soints have not changed. This modification provides additional assurance that failures or i

ma functions of equipment will not prevent the ability to recover and provide at least one train of cooling to both spent fuel pools. This modification does not introduce or increase the aotential for inadvertent loss of Spent Fuel Pool cooling. There are no potential failure modes ntroduced by implementation of the modification. The modification has no effect on accidents and malfunctions of equipment evaluated in Licensing Basis documents and does not create potential for a new type of unanalyzed event. This activity does not affect any system used for accident mitigation, will not affect the plant o ' response to a system failure, and will meet all system design basis requirements. There are r'o affected Technical Specifications for this activity. This activity has no adverse impact on the daily plant operations or outage activities.

All original design and safety criteria are met. Therefore, this modification does not involve an Unreviewed Safety Question.

_ to TXX 98021 Page 49 of 110 j

Evaluation Number SE 96 068 l

Revision 0 Unit: NN2 t

Activity

Title:

DM 96 022; LDCRs SA 96134 TR 96 002; Providing 480V Power Supply to Personnel Airlock Hydraulic Units Tripped By SI Signals; FSARITRM Revs Description of Changes:

A 480V feeder is provided to the personnel airlock from MCC 2EB12. Power at this MCC is de energized upon receipt of a Train A SI signal. A 480V shunt trip breaker, which de energizes the personnel altlock power feeder upon receipt of a Train B SI signal,is Installed in a )anelin the power feeder between MCC 2EB12 and the personnel airlock. An adm nistratively controlled 480V manual transfer switch is installed in the power feeder between the shunt trip breaker and alrlock hydraulle units which provides an alternate power source from the onsite 25kV loop during a Train A bus outage during Modes 5 & 6 only. The 25kV loop feed is protected by two breakers in series and included in the TRM to ensure surveillance requirements of Technical Specifications 4.8.4 (penetration protection) are met.

Also Licensing Basis Documents are updated to reflect the design and to justify separation requirements.

Summary of Evaluation:

The safety related purpose of this design modification is to remove power to the airlock automatically on the receipt of an SI signal. This is for ensuring that a postulated environmental failure of the airlock equipment during accident conditions are not violating containment integrity. Single failure proof design is provided by removing power with signals from both Sl trains Separation requirements within the shunt trip breaker panel are justified by l

failure analysis. Bypass of trip signals by transfer switch is precluded by restricting its use to Modes 5 and 6 except during a short test window under compensatory measures. Failure modes identified are analyzed and found to be either precluded by design or with alternative action available. Also basis for Technical Specifications are not impacted.

Based on the results of the safety evaluation, implementation of the activity do not involve an unreviewed safety question.

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i t) TXX 98021 Page 50 of 110 Evaluation Number SE. 97 001 Revision i Unit: 1N2 Activity

Title:

ONE Form 96 001555 Cort (ctive Actions, Revisions to Emergency Procedures for ECCS Transfer & Update of applicable FSAR Sections Description of Changes:

ONE Form 96 001555 describes differences within the FSAR and between the FSAR and plant calculations and emergency arocedures. As part of the corrective action, the FSAR will be updated to match the revisec emergency procedures and revised calculations. The applicable steps in the emergency procedures will be clarified in the FSAR consistent with the calculation of ECCS transfer volumes. ECCS transfer volumes will be updated. Instrumentation descriptions will be updated and clarified.

Summary of Evaluation:

No ne failure modes are introduced by these changes to the FSAR, procedures and calculations. The probability of reaching the empty setpoint prior to completion of ECCS transfer is not increased. The margin of safety as described in the BASES for the RWST Technical Specification volume is not reduced based on a re analysis of the required ECCS transfer volume.

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- to TXX 98021 L

Page 51 of 110 Evaluation Number SE 97 002 i

Revision 0 I

Unit: NN2

- Activity

Title:

Design Modification for Letdown Isolation on Safety injection Signal"S" for Valves 2 LCV 0459

& 2 LCV 0460 Description of Changes:

This Unit 2 design modification provides an automatic closure on a Safety injection Signal for Letdown isolation valves 2 LCV 0459 and 2 LCV 460. This automatic closure will then automatically cause the Letdown Orifice Isolation valves 2 8149 A, B, C to go closed due to the present electricalinterlock scheme. By providing this automatic isolation on a Safety injection Signal, relief valve 2 8117 will not lift (due to the closure of Containment isolation Valves 8152 and 8160 on safety injection actuation signals) eliminating an ALARA concern and ensuring manual operator action is not required during a safety injection signal.

A new handswitch will also be installed that will give the Operator the ability to bypass the Safety injection Signal for both valves 2 LCV-0459 and 2 LCV 460. This bypass will be used during the performance of surveillances when closure of these vtalves is not desired.

The systern design basis document and the FSAR (Table 7.3-4) will require updating to reflect 2-LCV 0459 and 2 LCV-460 as equipment actuated on a Safety injection Signal. They will appropriately note that the ESFAS function is "fJone" since the closure is not required for Nuclear Safety.

Summary of Evaluation:

Containment isolation Valves 8152 and 8160 close on safety injection actuation signals. If the letdown isolation valves LCV 0459 and LCV 460 are not closed, as has happened during spurious events, relief valve 8117 has lifted (as designed) and exceeded the capacity of the i

pressurizer relief tank. This ALARA concern was addressed by changes to the Emergency Operating Procedures (EOPs) to close the valves manually which is an " operations work around." By providing automatic isolation on a Safety injection Signal, relief valve 2 8117 will not lif t eliminating the ALARA concern and ensuring manual operator action is not required during a safety injection signal. This Safety injection actuation is solely for the purpose of assisting in the response to a Safety injection Signal so that the EOP step is not considered an operator work around. This EOP step will not be altered because it is acceptable to have the Operator verify that the automatic Letdown Isolation of the "S" signal did occur. This EOP step may be removed from the procedure in the future as it is no longer needed for a spurious St.

The function of these valves,(to prevent the uncovering of the pressurizer heaters on low pressurizer level), has not been altered. The added signalis not required for safety.

The new handswitch was evaluated for Human Factors and was found to be acceptable based on the guidance in NUREG 700. The addition of the Class 1E cable and switch, and the use of an existing Safety related relay will not create any event that has not been analyzed previously and will not alter or impede the function of these valves. Any cable or relay fault or open circuit will remove power to the valve supporting the close function of the valve.

. - _ _ _. _ _ _ _ _. _ _.._ to TXX 98021 i

Page 52 of 110 Evaluation Number SE 97 003 Revision 0 I

Unit: NN2 Activity

Title:

DM 96157; LDCRs SA 97-006, TR 97 004; Addition of Time Delays to All Four Overtemperature N16 Trip Channels for Unit 2 Description of Changes:

A time delay is added to each N16 Overtemperature trip channel to prevent unit trip from spurious signals, usually resulting from lightening strikes. The time delay is produced by adding a new card in each 7300 Protection cabinet. N16 Overtemperature trip response time requirement is increased to 8 seconds. Implementation will be at power, a Protection cabinet

.at a time, the channel being in trip during physicalimplementation, and in bypass during testing, as long as Tech Spec limit mandating channel be in trip is not exceeded. Licensing Basis documents, including Technical Requirements Manual, are updated to document design and time response increase.

Summary of Evaluation:

increased time response is justified by analysis Since Reactor Trip system is safety related, a single failure will not prevent completion of safety function required by accident analyses, and does not constitute a new failure mode. Setpoints and Allowable Values in Tech Specs remain valid with the increased response time.

All failure modes of the new components are detectable through either control board indication or during required surveillance testing.

Two accident analyses are affected by an extension of the OTN-16 response time; the uncontrolled rod withdrawal at power (FSAR Section 15.4.2), and the inadvertent opening of a

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pressurizer PORV or pressurizer safety valve (open PORV)(FSAR Section 15.6.1). These analyses were performed with NRC approved methodologies (minlmum DNBR criteria continue to be met).

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I Attachment i to TXX 98021 Page 53 of 110 Evaluation Number SE.97-004 Revision 0 l

Unit: INN Activity

Title:

DM 96156; LDCRs SA. 97 005,TR.97-003; Addition of Time delays to all 4 U1 Overtemperature N16 Trip Channels (Unit 1 change only) l Description of Changes:

A time delay is added to each N16 Overtemperature trip channel to prevent unit trip from i

spurious signals, usually resulting from lightening strikes. The time delay is produced by adding a new card in each 7300 Protection cabinet. N16 Overtemperature trip response tii requirement is Increased to 8 seconds. Implementation will be at power, a Protection cabint.

at a time, the channel being in trip during physicalimplementation, and in bypass during testing, as long as Tech Spec limit mandating channel be in it:p is not exceeded. Licensing Basis documents, including Technical Requirements Manual, are updated to document design i

and time response increase.

Summary of Evaluation:

Increased time response is justified by analysis. Since Reactor Trip system is safety related, a single failure will not prevent completion of safety function required by accident analyses, and does not constitute a new failure mode. Setpoints and Allowable Values in Tech Specs remain valid with the increased response time.

All failure modes of the new components are detectable through either control board indication or during required surveillance testing.

Two accident analyses are affected by an extension of the OTN 10 response time; the uncontrolled rod withdrawa: at power (FSAR Section 15.4.2), and the inadvertent opening of a prassurizer PORV or pressurizer safety valve (open PORV)(FSAR Section 15.6.1). These analyses were performed with NRC approved methodologies (minimum DNBR criteria continue to be met).

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___ _ _ _ to TXX 98021 Page 54 of 110 Evaluation Number SE. 97 006 Revision 0 Unit: IN2 Activity

Title:

LDCR SA 97 010; Addition of High Head Safety injection isolation Valves 8801 A & B to Equip.

List Not for Testing at Reactor Power Operation Description of Changes:

Adding High Head Safety injection Isolation valves 8801 A & B to the list of equipment that cannot be tested at reactor power so as not to damage equipment or upset plant operation as described in FSAR Section 7.1.2.5.

Summary of Evaluation:

The Inservice Valve Testing (IST) Plan, Table 13 Safety injection, Note 2 says the valves 8801 A & B cannot be full stroke exercised during plant operation because opening the valves results in unnecessary thermal transients on the Reactor Coolant system (RCS) cold leg nozzles for which they are not designed and imposes hydraulle transients on the charging system and on the Reactor Coolant Pump seals which can cause them to cock.

Opening the valves also results in an uncontrolled levelincrease in the pressurizer and an actuation of RCS pressure isolation check valves 8815, SI 8900A, B, C, & D which then requires leakage to be verified within its limit per Technical Specification (TS) surveillance 4.4.5.2.2.d. The affected unit must be placed in a shutdown condition (MODE 3,4, or 5) to support the leak test.

FSAR Section 7.3.2.2.5 says that testing of engineered safety actuation signals requires final devices to actuate. However, an end device that is actuated can cause plant upset / equipment damage, the end device will be declared inoperable and then disabled from operating by removing fuses / opening breakers, etc. FSAR 7.1.2.5 provides the list of exceptions for such end devices; however, valves 8801 A & B are not on the list.

This activity revised the FSAR to be more consistent with the contents of the CPSES IST Plan.

The IST Plan requires that these valves be tested every refueling because quarterly and tdd shudown testing is not viable. The reason this test is not viable is described in the par % graphs above. The salve relay tests for the control circuitry, however, did stroke these valv0s quarterly. This was achieved by running the positive Displacement Pump (PDP) for CVCS and isolating the Sl header. The IST Plan only takes into account the safety related pumps that are required for ECCS function when looking at the viability for performing the test. The PDP can be out of service for long periods of time with minimalImpact on plant operations. In fact,it is due to one of these extended OOS timer ' hat this situation was discovered. The IST Plan is correct and the 8801 valves should only oe stroked at the refueling frequency specified in the IST Plan. This allow the proper operation of CCPs without impacting power operation. Since this activity is to update the FSAR to be more consistent with what is already in the IST Plan and since the only change to the slave relay test is to remove the 8801 valves from operating, there are no new potential failure modes introduced by implementation of the proposed activity.

The containment isolation function is not affected since the valves remain closed during the surveillance tests. The open function on Si signal will be defeated during the surveillance tests, however it has no effect on accidents and malfunctions evaluated in the LBDs because the proposed activity is performed within window of inoperability allowed for testing per Technical Specifications. The proposed activity does not create a new type of unanalyzed event. There is no change to the acceptance limits or margin of safety. Therefore, this activity does not involve an Unreviewed Safety Question.

to TXX 98021 Page 55 of 110 Evaluation Number SE 97 007 Revision 0 Unit: IN2 Activity

Title:

ONE 95 0333; Emergency Diesel Generator starting air receiver safety to compressor class break Description of Changes:

DBD ME 028 requires that ANSI N 18.2/N18.2a be met for safety class 3 to NNS Interfaces.

NNS instruments connected via normally open manual valves to safety class 2/3 systems must be mounted selsmic category I and the instrument have a selsmic evaluation for pressure boundary. This is to meet the requirements of ANSI N18.2 as described in the FSAR In Section 3.2. Contrary to this requirement the EDG Starting Air compressors and instrumentation do not meet this seismic requirement. The instrument air lines run per DCN 6700 are not run in accordance with CPES I 1018 and FSAR Section 17A. TE-95 345 evaluated the tube run as notimpacting operability. TE 97 89 addresses the seismic capability of the EDG Starting Air Compressors. Compensatory measures are in place to provide additional assurance that the

- pressure boundary is not lost.

Summary of Evaluation:

The as built configuration of the EDG Starting Air Receiver to EDG Starting Air Compressor does not have a qualified interface between safety and NNS as required ANSI N 18.2/N18.2a.

The vibration of the compressors in normal service is above the seismic accelerations. It follows that the compressor pneumatic controls pressure boundary should not fall during a DBE. In addition compensatory measures are in alace to enhance the operational awareness and to decrease the possibility of common mode failure. Therefore the as built configuration of the compressor pneumatic controls does not result in a potential unreviewed safety question because the probability of a malfunction of the EDGs is not increased, the probability of a Station Blackout is not increared and the margin of safety in the basis for the Tech Specs is not reduced. An operability evaluation has been performed and the plant is operable. The non-conforming condition requires correction. Until this non conforming condition is removed, adequate compensatory measures will remain in place to offset the potential slight increased probability of a seismically induced failure.

Attachment i 13 TXX 98021 i

Page 56 of 110 Evaluation Number SE 97 009 Revision 0 Unit: NN2 i

Activity

Title:

Unit 2 Auxiliary Feedwater Terry Turbine Steam Supply and Drain Modification Description of Changes:

This modification to the steam supply lines and associated drains to the turbine driven auxillary feedwater pump turbine is intended to reduce overspeed trips and increase the reliability of the TDAFW Pump. The " tee" connection and upstream check and block valves for the two steam supply lines to the TDAFW pump turbine are being relocated from the pump room to the main steam penetration area to reduce the volume of pipe between the steam admission valves and the turbine. One of the existing supply lines will be retained as the cummon header and will have " knockout pots" installed to remove condensation generated by cold start transients. The other supply line will be converted to a drain / vent line for the knockout pots and will be routed to a new flash tank. The existing high and low pressure drains will be separated and routed to the flash tank, which will be vented to atmosphere and drain to the floor drain system. The turbine governor valve stem drain will be rotated to improve drainage. A drain valve will be installed on the turbine lube oil equalization line to ease draining of oil and provide assurance of moisture removal from the oil. The h!gh point leakoff line for service water inleakage detection is being rerouted and will no longer go directly to a floor drain. The flow restrictor in the line will be relocated to the end of the pipe. The pump sealleakoff and skid drains are being rerouted.

Summary of Evaluation:

This modification will reduce the volume of condensation generated and will remove the remaining water before it passes through the turbine, resulting in smoother starting transients ar.d a lesser demand on the governor to prevent overspeed trips. The flash tank will ensure that the steam from the drains and knockout pots does not create a harsh environment for the Class 1E equipment in the area. The mods will be in accordance with the applicable codes and standards per the FSAR. The overall effect will be improved reliability of the TDAFW pump.

There is no unreviewed safety question concerning these mods, d

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- Attachment 1 to TXX 98021 Page 57 of 110 Evaluation Number SE 97 010 Revision 0 Unit: 1N2 Activity

Title:

Delete Condensate Pump Trip on Loss of Main Feedwater Pump Oil Pressure Description of changes:

Deletes the trip of the Condensate pumps on low oil pressure on both Main Feedwater pump turbines.

Summary of Evaluation:

The purpose of tripping the condensate pumps on loss of oil to the Maln Feedwater Pump Turbine's (FWPT's)is to prevent 'windmilling' of the FWPT's with no oil to the bearings This trip is for equipment protection of the Main Feedwater pumps only and does not impact plant safety. The change will not affect the trip of the Main Feedwater pump turbines on loss of oil.

Neither the Main Feedwater nor the Condensate pumps are safety related, and they are not credited in accident analysis. Therefore, this change will not result in an unreviewed safety question i

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Attachment i to TXX 98021 Page 58 of 110 Evaluation Number SE 97 011 Revision 0 Unit: NN2 Activity

Title:

4 TM #2 97 001, 002,-003 & 010, provide temporary power to plant equipment during the i

implementation of DM #96 011 i

Description of Changes:

During the implementation of DM #96 011, temporary power is required to support the DM plan.

TM 2 97-001 provides temporary power to DC panels. 2 ED1 1 & 2ED12 utilizing the prototype 4

SCl spare battery charger and non 1E battery charger BC2D24.

TM 2 97-002 will give temporary power to DC panels 2ED21 & 2ED2 2 using the above same battery chargers, t

TM 2-97 003 will provide temporary 480 vac power (from 25 kv plant support power system -

j PSPS) to two (2) temporary battery chargers.

TM 2-97-010 willinstall temporary cable between JB1 A 5223 to non 1E or Class 1E battery.

Implementation of TM #2 97 003 & 2 97-010 and MM # 96-12 will glve temporary power which 4

will be used to perform maintenance testing on the Class 1E / non 1E batteries during the new battery charger's acceptance testing.

J Summary of Evaluation:

The TM activities will allow temporary power supplies (also changes temporarily plant drawings. FSAR,) for the duration of each Installed temporary modification during modes 5 & 6 only (except TM #2 97-003 which is for all modes) and required for implementation of DM # 96-011. During these modifications, one train will be affected and another train will be operable per TS. The TM are acceptable per plant design LBDs. The activity is consistent with the existing LBD and as such does not affect the probability or conseqbences of accidents or malfunctions of equipment important to safety already analyzed in the LBDs and does not affect TS, I

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- - - -. - to TXX 98021 Page 59 of 110 Evaluation Number SE.97-012 Revision 0 Unit: NN2 Activity

Title:

Modification of Control Logic for 2 FV 2239 to " Fall Closed" During Control Circuit Power i

Failure i

1 Description of Changes:

The controllogic for 2 FV 2239 is being changed to " fall closed"in the event of an electric i

power failure in the control circuit. The normal valve function, to open in order to maintain adequate flow and protect the Condensate Pumps, remains unchanged. Also, loss of air failure will continue to result in a " failed open" valve position.

Summary of Evaluation:

This change to the control circuit will result in the valve remaining in the normally closed position (normal power /high flow operations) or repositioning to a closed position (Iow power / low flow operation)in the event of a fuse failure. Operators will take other actions in the

- event of low flow coincident with a fuse failure to protect the Condensate Pumps, which have no safety function and are low energy pumps with sufficient time available for the operator to establish adequate recirculation flow. This modification results in more stable plant operations and continues to ensure the Condensate Pumps are protected from low flow conditions. A aositive effect will be realizer' for the accident analysis as the unit will be less susceptible to a oss of feedwater transient that can lead to a reactor trip.

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l to TXX 98021

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Page 60 of 110 Evaluation Number SE. 97 013 Revision 0 i

Unit: NN2 l

Activity

Title:

Administrative Control of Containment Isolation Valve 2 HV 3487 in an Inoperable Position for Operator Maintenance Description of Changes:

The internals of the air operator for Containment Isolation valve 2 HV-3487 are degraded and 4

must be replaced or repaired to avoid failure. The activity evaluated in this Safety Evaluation involves a one time procedure change notice to SOP 5098," Instrument Air System", to allow Containtnent isolation Valve 2 HV-3487 to be JNOPERABLE while continuing to supply instrument air to the users within containment. In order to maintain the Instrument air supply, a bypass will be provided around valve 2 HV 3487. This bypass will be installed using a rubber hose suitable for the instrument air system conditions connected to class 2 valve 201-0033 and which will ensure that containment isolation can be re established, if required. g this acti non safety valve 2Cl.0043. Compensatory operator actions will be taken durin Summary of Evaluation:

The activity associated with this one time procedure change notice to SOP 509B," Instrument Air System", has been evaluated for all potential credible failure modas. Compensatory operator actions will be taken during this activity to ensuie that containment isolation can be re-

- established. There are no credible failure modes or credible accidents or malfunctions of equipment important to safety that are adversely affected by this activity. The activity does not i

involve an unreviewed safety question.

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' to TXX 98021 Page 61 of 110 Evaluation Number SE 97-016 Revision 0 Unit: 1N2 Activity

Title:

DM 97-13 (Unit 1 & 97-15 (Unit 2))" Auxiliary Feedwater Steam Admicsion Valve Protective Train Swap" Dest.rlption of Changes:

This modification is being performed to resolve a non conforming condition that was identified in ONE Form 97-000074 in accordance with TU Electric's commitment in LER 97-001. Unit 1 and 2 Steam Admission Valves 1/2 HV-2452-1,2 protective train function needs to be

" swapped"In response to ONE Form 97 000074. This "svap" will place the A train valve on Main Steam Loop 4 and the B train valve on Loop 1. (This will require rerouting the power control cabling for each va!ve to it's opposite Train valve). This modification will resolve a i

condition where a single failure in one Train coincident with a feedwater line break in the other i

train would not allow the available steam supply to the steam driven auxillary feedwater pump. -

Summary of Evaluation:

The installation of this modification will reestablish the existing accident analysis in the Licensing Basis Documents. The FSAR and Technical Requirements Manual will be updated to elaborate the critical design features and Failure and Effect Modes Analysis in Section 10.4.9 and to update the respective figures / details. This modification brings the physical plant in agreement with the accident analysis previously documented in the FSAR. The existing testing requirements are unchanged by aligning the protective train valve to the appropriate steam line to meet the existing licensing and design commitments,

Attachment i t)TXX-98021 Page 62 of 110 4

Evaluation Number SE-97-017 Revision 0 Unit: 1NN -

Activity

Title:

Small break LOCA Analysis for Unit 1, Cycle 6.

Description of Changes:

Replace the Westinghouse NOTRUMP small break LOCA analysis described in FSAR Appendix

-4A. The new analysis is for Unit 1, Cycle 6 and is based on the TU Electric methodology approved by the NRC and described in detallin topical report RXE 95-001-P-A. Also update FSAR Chapter 15.6 to make the TU Electric small break LOCA analysis methodology the methodology for the small break LOCA analysis of record for Unit 1. The Westinghouse NOTRUMP small break LOCA analysis remains the analysis of record for Unit 2 until a cycle specific analysis is performed using the TU Electric methodology.

Summary of Evaluation:

The Unit 1, Cycle 6 small break LOCA was analyzed using the TU Electric methodology approved by the NRC and described in detallin topi:al report RXE 95-001-P-A. This topical report is included in the list of approved methodologies in T.S.6.9.1.6b item 20. All event acceptance criteria were found to be satisfied. Therefore, there is no unreviewed safety question in replacing the existing analysis of record, which was performed by Westinghouse using NOTRUMP.

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- to TXX-98021 Page 63 of 110 Evaluation Number SS-97-018 Revision 0 Unit: NN2 Activity

Title:

Containment Penetration modified for use in depressurization of containment during ILRT &

cable feed thru during outage Description of Changes:

Containment isolation penetration number 2-M111-017 is to be modified to allow for its use for depressurization of containment during ILRT and for cable feed thru during normal outage activilles.

Summary c' Fvaluation:

The exist' p ".nfiguration consists of a welded cap inside Containment, This cap is to be removed & c v modificati q of the penetration. The permar,ent installation will consist of a -

flange and L;.0d flan M Insioe Containment and a 90 degree elbow,12" pump piece, flange, bilnd flange, and 3/4" test connection (off the 12" pup piece) Outside Containment. This configuration conforms to 10CFR50 App. A, General Design Criteria 56,

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- to TXX-98021 Page 64 o' 110 Evaluation Number SE-97-019 Revision 1 Unit: 1N2 Activity

Title:

Allow an alternate method to cool the MS & FW penetration areas during extreme summer weather conditions (DCN 10757, R1)

Description of Changes:-

This activity revises DBDs ME-302,302C & 009 to allow an alternate method of operating the HVAC systems during extreme summer weather conditions. It proposes to open the doors of the MS & FW ventilation air-handling unit upstream of the cooling coil to provide an alternate path for intake air in lieu of using 100% outside air. Selective tornado dampers would also be placed in the open position to provide a flow path from the three Safeguards elevations to the intake of the MS & FW air handling unit, in order to ensure sufficient flow through the tornado dampers and into the air handling unit, the exhaust fans for the EAVS should be turned off.

The differential pressure of the Safeguard Electrical Area (SGEA) should be maintained slightly positive relative to the Primary Plant negative pressure boundary by turning off one or two of the EAVS oxhaust fans as appropriate. Tornado dampers located in the stairwell should be opened slightly to ensure that any differential pressure due to the new method of operating the HVAC systems will not preclude the stairwell fire doors to the stairwell from closing and will equalize pressure at the different building elevations, By opening the air-handling unit upstream doors and the tornado dampers, a MS & FW area supply air path of less resistance is opened. Instead of utilizing 100% outside air, a mixture of outside air and tempered air from the SB Electrical Area will be used as supply air to the MS & FW penetration areas. This addition of tempered air will allow the air mixture of outside air and tempered air from the SB electrical area will be used as supply air to the MS & FW penetration areas. This addition of tempered air will allaw the air mixture going through the MS & FW air-handling units, to be cooled to a lower temperature than if the unit would draw 100% outside air.

Summary of Evaluation:

Revising the iespective DBDs to allow this alternate method of operation during extreme summer weather conditions does not ccastitute an Unreviewed Safety Question. Opening the tornado dampers discussed above does not increase the heat load significantly to affect the performance of the safety related fan coil units if there were an accident while operating in this lina-up. The open tornado dampers will help equalize building pressure differentials were a tornado to occur while operating in this line up. The components needed to protect the Train A

& B electrical areas from steam and feedwater line cracPs or smallline breaks in the penetration room are unaffected by this activity. The Margin of Safety is unchanged.

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- to TXX 98021 Page 65 of 110 Evaluation Number SE-97 020 Revision 0 -

Unit: 1N2 Activity

Title:

LDCR SA 97-035; Polar Crane Operation During MODES ithru 4 to Support Polar Crane inspection, Functional Checks and Prev. Maint.

l Description of Changes:

Prior to entering outages, polar crane operation is required to perform certain inspections, functional checks and preventative maintenance (PM) activities. To preclude the crane from being a restraint to refueling operations the PMs should be performed prior to the refueling outages. The polar crane may be energized during MODES 1-4. The crane will remain essentially In the park position (FSAR Figure 1.2-8). No load will be carried during MODES 1-

4. A compensatory measure (operator, in direct communication with the control room, at the power supply disconnect outside containment) is established to prevent inadvertent operation of crane due to High Energy Line Break in containment.

Summary of Evaluation:

Should a LOCA or MSLB occur auring crane operation the potential exists for the crane to operate spuriously due to environmental effectc causing damage to the safety related SSCs inside the reactor building. To preclude this from occurring during operation of the crane during MODES 1-4 an operator in communication with the Control Room shall be stationed at the power supply disconnect outside containment. Should either area become uninhabitable the operator shall open the power disconnect and leave the area. Any movement of the crane should be small enough so that the operator can safely exit the crane. The crane is to remain essentially in the designated park position (FSAR Figure 1.2-8). The hook should not be placed in a location to impact any safety related structure system or component ( e.g. hydrogen recombiners). No load is to be carried by the crane. The limitations imposed by this change will be incorporated into the appropriate maintenance procedures. This activity does not pose any unreviewed safety questions.

Attachment i to TXX 98021 Page 66 of 110

' Evaluation Number SE. 97 021 Revision 0 i

Unit: 1X2 Activity

Title:

Remove conductivity monitoring elements from the Condensate System in both units.

Description of Changes:

The modification removed conductivity elements from each Unit which monitored lake water in-leakage into the Condensate System. Two elements were located in the outlets of the two Auxillary Condensers and the other 16 were in the Main Condenser hotwells. The elements never worked as intended ard also presented air in-leakage paths into the Condensate System. Associated electro.11cs and instrumentation also were removed.

Summary of Evaluation:

This chan0e to the plant will have no negative affect on the Secondary Side Chemistry Program. Condenser water in-leakage into the Condensate System is still monitored by other means. Further, physically removing the Instrumentation and plugging the access points eliminates air in leakage points which have caused problems in the past. It should be noted

that the elements were actually taken out of service in 1994 and this activity only removes them from the plant.

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Attachtnent 1 to TXX-98021 Page 67 of 110 Evaluation Number SE.97-022 Revision 0 Unit: 1X2 Activity

Title:

Loss of Control Room Annunciation of Component Cooling Water Low Flow Out of the Containment Spray Heat Exchangers Description of Changes:

Final Safety Analysis Report (FSAR) Paragraph 9.2.2.3 states that low CCW flows are annunciated in the Control Room. Control Room alarm of low flow of CCW out of the CT heat exchangers was lost inadvertently with implementation of Design Modifications93-042 and 93-043 on Units 1 and 2, respectively. This evaluation determines the impact to safety, pending -

restoration of the alarms per d:sposition of Operations Notification Evaluation (ONE) Form 96-137.

i Summary of Evaluation:

The Control Room alarm annunciation circuits for CCW low flow out of the CT heat exchanges are not safety related (Non-1E) and are not required as accident monitoring instrumentation.

-_ Control Room Operators have available other credited means of determining critical valve positions, flows, and temperatures related to evaluating heat transfer performance from the CT heat exchangers. It is concluded therefore that the activity does not create or increase the probability or consequences of equipment malfunctions or accidents affecting safety, and does not decrease the margin of safety as defined in the basis of any. Technical Specification,

ttachment i to TXX 98031 Page 68 of 110 Evaluation Number SE-97-024 Revision 0 Unit: 1NN Activity

Title:

Manual Makeup to the Reactor Coolant Pump (RCP) Seal #3 Standpipes and the Pressurizer Relief Tank (PRT)

Description of Changes:

4 Operations Notification Evaluation (ONE) Form 96-1473 documents pressure transients which frequently result in inadvertent lifting of thermal relief valves (RC-0036, DD-0600) outside containment along penetration Mill-1 in Unit 1. This same condition existed in Unit 2 previous to implementation of a Design Modification and was thought to have been resolved following implementation of a Design Modification in Unit 1. Although Mill-1 is a non-essential penetration (Reactor Makeup Water to PRT/RCP Seal Standpipes), the location of one of the thermal reliefs between containment isolation valves requires further analysis of this condition and actions taken by Operations to lessen the impact of this condition on system operation.

Makeup to the RCP Seal #3 Standpipes is being conducted as a manual operation as an interim corrective action until the pressure transient problem is resolved, This is accomplished by keeping valve 1-8047 (Reactor Makeup Water to Pressurizer Relief Tank 1-01/ Containment

. Supply Outside Reactor Containment le stion Valve) normally closed and opening at once a n this is a temporary condition, it is not the shift to auto-fill to the standpipes. Alth-s configuration depicted in the Licensing Basis Documents.

Summary of Evaiuation:

Manual operation of the isolt i.1 valve 1-8047, Reactor Makeup Water to the RCP Seal #3 Standpipes and the PRT dos..iot impact the ability to makeup to these components in a timely manner. This is not an essential penetration during off-normal plant conditions. There is no increase in the probability of a Licensing Basis Accident or safety related equipment malfunction nor is there a potential for creating a previously unanalyzed event. There is no impact on Technical Specifications and therefore there is no potentialimpact on any margin of safety.

-_ to TXX-98021 Page 69 of 110 Evaluation Number SE.97-026 Revision 0 Unit: 1X2 Activity

Title:

LDCR PC-95-01; Revise and Relocate the CPSE" Process Control Program into the Station Administrative Manual (STA-713)

Description of Changes:

This activity revised the CPSES Process Control Program (PCP), Revision 4, and converted the document into the Station Administrative Manual as procedure STA-713," Process Control Program",

This activity was administrative in nature and involved revismg and updating the existing approved program which ensures that CPSES processes and packages wet solid radioactive waste to meet regulatory requirements for disposal. This change revised the program to meet the format of the Station Administrative Manual. Additional revisions / updates were incorporated to better reflect current CPSES wet waste processing practices and methods, delete Dry Active Waste (DAW) from the program scope, revise program discussion concerning Mixed Waste and major changes to liquid, gaseous and solid radwaste treatment systems which'are covered by other station programs (e.g.,50.59 process snd Mixed Waste Control Program per procedure STA-633), update program responsibility per the current CPSES organization, and expand requirements and discussion for processing radioactive sludge or miscellaneous liquids.

Summary of Evaluation:

The PCP as revised and incorporated in the STA procedure retains the required administrative and program control elements to ensure that the subject final radwaste product and its shipment to and acceptance for burial meet state and federal requirements. These changes do not affect plant safety related equipment, systems or components utilized for radioactive waste processing. Existing failure modes are bounded by the Liquid Waste System teak or failure. No new processing activities are associated with this revision; accordingly, it was determined that this activity does not involve an unreviewed safety question.

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1 to TXX 98021 Page 70 of 110 Evaluation Number SE-97-027 Revision 0 Unit: NN2 Activity

Title:

Addition of CP2 EIPRLV-48 to Multi Train Equipment Lists; DCN-10781 R0 and LDCR SA 97-l 036 Description of Changes:

CP2 EIPRLV-48. identified as multi-train equipment and iniernal separation per RG 1.75 is 2

not required for this panel. This activity adds CP2-EIPRLV-48 to the list of electrical equipment not requiring internal cable separation to FSAR T8.3-10 and adds exception in 1 A(B) and 8.3.1.2.1 to justify exemption from internal train separation requirements of RG 1.75 (FF#183).

Summary of Evaluation:

This change identifies CP2 ElPRLV-48 as multi train equipment that is exempt from Internal 3

train separation requirements. CP2-ElPRLV-48 is the U2 TDAFWP control panel. The panel does not serve a safety related function and contains one non safety related cable along with several Train A trouble alarm and the tachometer. Power to the associated cables is tripped upon receipt of an SI signal. The maximum credible fault current in the non-safety related cable is less then the current carrying capacity of the cable and will not affect any of the associated cables in its vicinity (even touching). Therefore, the separation between non-safety and associated train A calles in CP2-E!PRLV-48 is not required.

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to TXX-98021 Page 71 of 110 Evaluation Number SE-97 028 Revision 0 Unit: NN2 Activity

Title:

TM 2-97-007 - Provide temporary 480 VAC power to Battery Charger BC2D2 during Train 'B" outage in 2RF03 Description of Changes:

Provide temporary 480 VAC power to Battery Charger BC2D4 during 2RF03 Train "B" BUS outage to maintain power supply to 125 V/250 VDC loads and to maintain the battery in a fully charged condition. This Temporary Modification will only be in place during Train "B" BUS outage and only during mode 6.

Summary of Evaluailon:

The use of an existing plant support feed to battery BC2D4 to temporarily supply BC2D2 in parallel when the battery is fully charged and the bus loads are minimal is acceptable for the duration of Train "B" outage during mode 6 in 2RF03. The FSAR discussion & figures and DBD EE 044 & 057 are impacted temporarily by this TM bt.. It will not create any new failure modes. TM will be in compliance with RG 1.75. System interaction, Fire Safe shutdown, and commodity clearance issues are not applicable during mode 6. The plant support power feed has been previously evaluated as acceptable for use during outages. The potential failure modes for the cable and connection is the same as the existing installation (i.e., faults, open circuit, and cross connection), which has been previously analyzed. Therefore, this TM does not involve an Unreviewed Safety Question. As long as separation is maintained per DBD-EE-057, there will be no Safety impact since the Non-1E battery chargers are not required for safe shutdown.

. to TXX 98021 Page 72 of 110 Evaluation Number SE-97-032 Revision 0 Unit: 1N2 Activity

Title:

SOP 7.P-978-2," Bladder Evacuation of Condensate Storage Tank" to facilitate nitrogen removal from under the diaphragm on the CST Descrliition of Changes:-

This activity involves a temporary procedure that is utilized to draw a slight vacuum on the Condensate Storage Tank (CST) above the tank diaphragm in order to lift the diaphragm away from the surface of the water and the wall of the tank. This willimprove the ability of an installed vacuuri pump to draw off nitrogen that has collected under the diaphragm. A water seal is u'ilized at the CST vent connections to allow the vacuum to be drawn on the tank but limit the omount of vacuum that can be achieved.

1 Summary of Evaluation:

The installation of these components has been evaluated and it nas been shown that the Auxillary Feedwater and Condensate Storage Tank will be capable of performing their safety functions during this activity. The margin of safety as required by the FSAR and the design code for the tank is not reduced by this activity. Therefore the system and components remain OPERABLE during this activity. This activity willimprove the ability to remove nitrogen that has collected under the CST diaphragm. Consideration has been given for all the potential failure modes for this activity and it has been determined that there are ao credible failure modes that could adversoly affect systems, structures or components resulting in an increase in the probability, severity, or consequences of any accident analyzed in licensing basis documents. The activity does not irvolve an unreviewed safety question or require an amendment to the Technical Specifications. This activity does not adversely affect the Auxiliary Feedwater System, the CST or any other system used for accident mitigation and will not impact plant response to an upset, emergency or faulted condition.

4 1

. to TXX 98021 Page 73 of 110 Evaluation Humber SE-97-035 Revision 0 Unit: 1N2 i

Activity

Title:

DCN 10519, Rev.0; ONE-96-721; LDCRs SA 96-131, IT-97-006, TR-97-007: Qualify MS-0101

& MS 0128 to function as containment isolation Description of Changes:

Main Steam isolation valves which supply the Auxiliary Feedwater turbine require manual closure capability, The intent was to add hand wheels to the existing AOVs (HV-2452-1,-2)in Order to provide for manual closure of the valves following exhaustion of air accumulators. This option is not economically feasible therefore the upstream block valves (MS-0101, -0128) must be qualified to function as Containment Isolation valves. This activity requires updates to the Design Basis Documents (DCN 10519, Rev 0, ONE-96-721), inservice Testing Program (LDCR-lT-97-006), FSAR (LDCR SA-96-131), and Technical Requirements Manual (LDCR-TR-97-007).

Summary of Evaluation:

Upstream block valves are capable of providing manualisolation of the steam supply lines to the Auxillary Feedwater Pump turbine. Manual closure of the block valses is only required following exhaustion of the air accumulators for HV-2452-1,-2 after loss, of instrument air.

Since these valves are (1) ASME Code Class 2 manual gear-operated gate valves, (2) accessible for operator action, and (3) will be added to the IST Plan, contsinment isolation is ensured during off-normal plant conditions. There is no increase in the probability of a Licensing Basis Accident or safety-related equipment malfunction nor is there a potential for creating a previously unanalyzed event. There is no impact on Technical Specifications and therefore there,s no potentialimpact on any margin of safety.

. to TXX-98021 Page 74 of 110 Evaluation Number SE-97-036 Revision 0 Unit: 1X2 Activity

Title:

Delete Procurement Engineering & Replace with VETIP in the "CPSES Response" Section Description of Changes:

Change FSAR Section i.C.5 to identify that Vendor Technical Correspondence is reviewed for -

Operating Experience as part of the VETIP Program.

Summary of Evaluation:

The change to this section of the FSAR is to clarify that Operating Experience is screened as part of the VETIP Program raher than nam *.ig a specific group within the program. The screening of Operating Expenance has been and will continue to be part of the VETIP program.

This change has no adverse effect on structures, systems, or components, nor on equipment malfunctions, postulated accidents, or Technical Specifications. This change is based on the CPSES response regarding the screening of vendor technical correspondence for Operating Experience and does not ruult in a reduction of plant commitments.

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_. _ 4 TXX-98021 Page 75 of 110 Evaluation Number SE 97 037 Revision 0 Unit: NN2 Activity

Title:

DCN 11199 R/0 LDCR SA 97-049. Restores remote current monitoring function at main contro!

board for the RCP Motor 2 04 Description of Changes:

This evaluation is applicable to DCN-11199 R/0 and LDCR SA 97-049. DCN-11199 provides engineering details and justification to restore remote current monitoring function for reactor coolant pump 2-04. LDCR SA 97-049 revises FSAR figure 8.3-5 Sh. 2 to show the restoration.

NOTE: The remote current monitoring function was previously disabled by a faulted signal cable. The faulted cable was disconnected thereby rendering the current monitoring function inoperable. This activity restores the circuit to its original design which was part of the original licensing basis of CPSES U2. The previous disabling of this function was approved under ONE 96-080, DCN-10507, LDCR SA 96 090 and SE 96-059.

i I

This activity effectively reverses changes made under DCN-10507 and LDCR SA 96-099.

Summary of Evaluation:.

DCN-11199 R/0 is acceptable per current licensing and design basis commitments and requirements. The changes made do not affect any safety related equipment and h ve no adverse impact on any other plant equipment. This change restores a diverse means of monitoring the condition of the reactor coolant pump motor 2-04 and thereby facilitates and enhances operation of the reactor coolant system. DCN-11199 R/0 restores the current monitoring circuit for RCP 2-04 to its original configuration.

1

- - _ - to TXX-98021 Page 76 of 110 Evaluation Number SE 97 038 Revision 0 Unit: NN2 Activity

Title:

Water Hammer Issues Related to Unit 2 Heater Drain System Description of Changes:

The Unit 2 heater drain system is modified to reduce water hammer events. The Heater Drain Pump recirculation lines are extended inside the Heater Drains Tanks to below water level.

The pump discharge and alternate drain valves are resized and their isolation valves are replaced. The Heater Drain Pump isolation valves and discharge check valves are replaced.

Warm up lines are added where appropriate. New nozzle check valves are added in the drain t

lines of Feedwater Heaters 2A and 28.

Summary of Evaluation:

The modifications will reduce the frequency and severity of water hammer events in the Heater Drains System. All the systems involved in the modification are not required for safety, have no protective functions, and can not impair the ability of protection systems to function. The.

only accidents in Chapter 15 that could be impacted by the SSCs associated with this proposed

- activity are those described in Chapter 15.1, " Increase in Heat Removal by the Secondary System" and Chapter 15.2, " Decrease in Heat Removal by the Secondary Systems". The probability of occurrence of these accidents is expected to decrease due to this activity. There is no unreviewed safety question.

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- Attachment 1 to TXX 98021 I

- Page 77 of 110 Evaluation Number SE-97-039 Revision 2 Unit: 1X2 Activity

Title:

10kVA Class 1E Elgar inverters and 7.5 kVA NSSS Westinghouse inverters w/10 kVA Class 1E SCI inverters (b) 10 kVA Non Class 1E Elgar inverter Description of Changes:

The DM activities are as follows: a) Replacement of 10 kVA Class 1E Elgar inverters with 10 kVA Class 1E SCl inverters b) Replacement of 10 kVA Non-Class 1E Elgar inverters with 10 kVA Non Class 1E SCIinverters. c) Replacement of 7.5 kVA NSSS Westinghouse inverters with 10 kVA Class 1E SCIinverters. d) Replacement of 225 Amp Class 1E PCP battery chargers with 300 Amp Class 1E SCI battery chargers. e) Addition of Class 1E 10 kVA SCI inverters (and associated cables) as an installed spare per train, f) Deletion of Class 1E,480 1

V AC feed to the inverters from MCCs 2EB1-1,2EB2-1,2EB3-1 and 2EB4-1. g) Deletion of Non-Class 1E,480 V AC feed from MCC 2B2-3 to the bypass transformer T2C3 and rewire Inverter 1V2C3 feed from Class 1E MCC 2EB3-1 to the bypass transformer T2C3. h) Deletion of Non Class 1E feed from MCC 2EB1-3 to Non-Class 1E inverter 1V2C2.1) change of inverters circuit breaker ratings to 150 Amp in Class 1E DC Switchboards 2ED1,2ED2,2ED3, 2ED4 and Non-Class 1E, DC Switchboard 2D2. j) Change of battery chargers circuit breaker ratings to 500 Amp in the Class 1E DC Switchboards 2ED1,2ED2,2ED3, and 2ED4. k)

Furnishing of 120 V AC feed to the bypass circuit of Protection Channel and installed spare inverters from Class 1E 120 V AC Panels 2EC3 and 2EC4.1) Addition of bypass switches to provide alternate power supply for Non-Class 1E 118 V AC Panels 2C2 and 2C3.

Summary of Evaluation:

4 The impact of DM activities on Fire Safe Shutdown Analysis (FSSA), Fire Protection, Seismic Interaction, System Interaction, Tornado Venting, Plant Safe Shutdown, Emergency Diesel Generators (EDGs) loading, Electrical Protection, Electrical separation / isolation, Acceptance of Electrical Equipment Ratings and Adequacy of Voltage for Device Operation was reviewed. It was determined that there are no credible failure modes associated with the implementation of DM 96 011 activities.

The impact of greater heat losses with the new equipment was reviewed and it was determined that the slight increase in heat loads is within the capabilities of the Uninterrupted Power Supply (UPS) Heating Ventilation Air conditioning System. Therefore, the qualified life and operation of the equipment in the UPS rooms are not affected during all modes of plant operation (Normal, Loss of Offsite Power and Station Blackout Conditions).

The compensatory actions requiring the temporary sealing of the breached Equipment Qualification (EO) barriers as discussed in the DM implementation plan were reviewed. It was t

determined that these actions are acceptable and the CPSES EQ Barrier Program is satisf;ed.

Based on the above evaluation, it is concluded that the implementation of DM 96-011 does not result in an unreviewed safety question.

- - -. to TXX-98021 Page 78 of 110 Evaluation Number SE-97-040 Revision 9 i

Unit: 1N2 Activity

Title:

Evaluation of mode 1-4 usage of entmt bldg. elevators & the elec, of the equipment hatch winches for pre-outage prev. maintenance Description of Changes:

The use of the Containment Bldg. Elevators (CP1/CP2 MEELRB-01)is required to both support the one time initialInstallation of the equipment hatch winch for DM 95-084 by lifting heavy components from EL 83 l' to EL 860' and to facilitate personnel movement during at power containment entries. Abo, to support inspections, functional tests, preventative maintenance (PM) activities, and penodic load tests prior to initial use in a refueling outage, electric power is required to operate the winch (CP1/CP2-MEMHCH-41).

Summary of Evaluation:

Should a LOCA or MSLB occur while either the elevator or the winch are energized, the e

potential exists for inadvertent operation through failures in their controls as a result of environmental effects. Inadvertent actuation of the winch if attached to the equipment hatch may cause damage to safety related SSC inside the reactor building. (Failure of the conta;nment penetrations resulting from the energized circuits arcing or shorting to ground is not credible duc to the double breaker protection provided and the functionality requirements from the (Technical Requirements Manual.) To preclude a failure associated with inadvertent or uncontrolled actuation of the winch from occurring during modes 1-4, the winch will not be energized for PMs while attached to the equipment hatch and a dedicated operator in communication with the Control Room shall be stationed at the p >wer supply disconnect breakers outside containment. Should either the area with the winch operators or the area of

'h3 power disconnect breakers become uninhabitable, the operator shall open the power disconnect and leave the area. Otherwise, during modes 1-4, the power supply breakers to the 480 VAC welding receptacles and the winches shall be normally OFF. The power supply breakers for the containment elevators may be left in the ON position durhg all modes of plant operation. The elevators are fully enclosed in a dedicated structure and do not affect any SSC by their movement or position within the elevator shaft. The limitations in this SE will be Incorporated into the appropriate Operations and Maintenance Procedures.

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l to TXX-98021 i

Page 79 of 110 Evaluation Number SE-97-041 Revision 0 Unit: 1N2 Activity

Title:

MM 96-012: Impair EQ Barrier penetrations to facilitate implem. of a permanent battery load -

test connection.

Description of Changes:

This activity involves the temporary impairment / breach of two EQ barrier penetrations located in the floor of corridor room X-122 in order to facilitate implementation of a permanent battery load test connection. A breach of these barriers may introduce an environmental flowpath between he UPS/ Distribution corridor room X-122 and the Air Compressor & Steam Generator Blowdown Heat Exchanger room X-113 contains SGBD and Auxiliary Steam system piping which have postulated high energy line breaks (HELB) that could subject safety related equipment in room X-122 and communicating rooms to harsh environmental effects, implementation of restrictions, limitations and comp' ~satory measures while breaching these barriers will satisfy the environmental concerns be ause the parameters of temperature, i

pressure, relative humidity and flooding will be malittained within acceptable limits should the described break occur.

Summary of Evaluation:

Consideration has been given for all potential failure modes for this activity and it has been determined that by implementing compensatory measures, no credible failure modes exist that could aversely affect systems, structures, or components resulting in an increase in the probability, severity or consequences of any accident analyzed in the LBDs. This act!vity does not adversely affect any system used for accident mitigation and will not impact plant response to an upset, emergency or faulted condition

Attachment i to TXXo98021 Page 80 of 110 Evaluation Number SE-97-043 Revision 0 Unit: NN2 Activity

Title:

Evaluation of the Effects of Extended AFD Target Bands on Unit 2 Cycle 3 Operation Description of Changes:

Appendix 4B of the CPSES updated Final Safety Analysis Report and the Core Operating Limits Report provide the values of the axial flux difference (AFD) target band as (+3,-12) %

AFD about the target in order to allow Unit 2 Cycle 3 to return to full power following a power reduction to approximately 50% RATED THERMAL POWER, the AFD target bands are extended to (+5, -12)% AFD about the target. The extended target bands result in a reference axial power shape, used in the DNB analysis, which is more limiting that the previous Unit 2, Cycle 3 reference axial power shape.

Summary of Evaluation:

The effect of the revised reference axial power distribution is offset by the DNBR benefits obtained through the use of a CPSES specific model of the effects of the lower plenum flow anomaly. Previously, a very conservative, generic model had been used. With the CPSES-specific model and the revised reference axial power distribution, all applicable event acceptance criteria have been demonstrated to remain satisfied; therefore, there is no change to the consequences of the accident as previously analyzed nor is there any reduction in the margin of safety as defined by the plant Technical Specifications.

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.... to TXX-98021 Page 81 of 110 Evaluation Number SE-97-044 Revision i Unit: NN2 Activity

Title:

CPSES 2, CYCLE 4 CORE CONFIGURATION Description of Changes:

During the third refueling outage for CPSES Unit 2, prior to operation of Cycle 4, seventy-six fresh Region 6 fuel assemblies and one once-burned Region 2 assembly will replace 28 Region 4B assemblies,40 Region 4A assemblies, and 9 Region 3 assemblies. For the CPSES Unit 2 Cycle 4 core configuration, seventy six fresh fuel assemblies manufactured by Siemens Power Corporation (SPC) will be co-resident with 96 partially burned fuel assemblies manufactured by SPC and twenty-one partially burned optimized fuel assemblies (OFA) manufactured by Westinghouse. A fresh Siemens fuel assembly (FF77)is replacing assembly FF63 which was damaged during fuel movement in the fuel building. This new assembly, FF77, will be loaded into the position originally designated for FF18 and FF18 will be loaded into the position origir; ally designated for FF63, the damaged assembly. Because FF63, FF18 and FF77 are essentially identical, the statements and conclusions in this revision of SE 97-0044 are the same as those in Revision 0 of SE 97-0044. Figure 2.2-1 is changed to correctly identify the assemblies in each core location and a statement regarding the FF18 and FF77 location swap is made in the summary section 1.3.

Summary of Evaluation:

This mixed core configuration has been evaluated for mechanical and thermal-hydraulic compatibility between the SPC and Westinghouse fuel assemblies. All applicable design criteria were determined to be satisfied. The neutronic characteristics of the Cycle 4 core configuration have been evaluated for their effect on the accident analyses. In all cases, it was determined that the applicable event acceptance criteria are satisfied. Because all mechanical design criteria continue to be satisfied, there is no reduction in any failure point introduced by the Cycle 4 core configuration. All acceptance criteria of the accident analyses continue to be satisfied; therefore, there is no increase in the consequences of any accident previously analyzed. Based on the foregoing, it is concluded that the Cycle 4 core configuration does not reduce any margin of safety as defined by the plant Technical Specifications; therefore, the proposed change does not involve an unreviewed safety question.

Attachment i 13 TXX 98021 Page 82 of 110-Evaluation Number SE 97-048 -

Revision 0 Unit: 1X2 Activity

Title:

Replace Spectacle Flanges with 4 Cross Connect Valves in the Component Cooling Water (CCW) Cross-tle lines for the SFP Heat Exchangers Description of Changes:

This activity restores the CCW Unit 1/ Unit 2 manual cross-connect valves and removes the spectacle flanges currently Installed. Spectacle flanges were installed during Unit 2 construction to provide positive leak-tightness. Previously, butterfly valves had been In service but were not adequately maintaining the leak-tightness required for this application.

Summary of Evaluation:

The function of the dual valves in the Component Cooling Water cross connects is to provide some additional measure of redundancy between Unit 1 and Unit 2 CCW systems when both Units are operating. During Unit 2 construction, the cross connect were replaced with spectacle flanges to ensure no cross-unit leakage. Restoring the cross connect with valves designed to Class VI leak tightness will ensure accurate maintenance of CCW inventory while enhancing the capability of providing a cross-unit supply in a timely manner.

There is no change in the probability of a Licensing Basis Accident or safety related equipment malfunction nor is there a potential for creating a previously unanalyzed event. There is no impact on any Technical specification and therefore there is no potential impact on any margin of safety.

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_ -. to TXX-98021 Page 83 of 110 Evaluation Number SE-97-049 Revision 0 Unit: 1NN Activity

Title:

Temporary Modification 1-97-003 Gagging of Valve 1-LV-2592 in Partially Open Position to Facilitate Repair of Valve Actuator

- Description of Changes:

TM 1-97 003 will gag valve 1-LV-2592 in a partla ly open position. Valve 1-LV-2592 is not responding to the level switches associated with the tanks and it is believed that the actuator is in need of repair. Valve 1-LV-2592 normally modulates to maintain levelin the Heater Drain Tanks, and small deviations in level can be made up by manually adjusting Steam Generator Blowdown Flow.

Summary of Evaluation:

The Heater Drain System is non safety related and is not required for safe shutdown of the plant. There are no accidents or malfunctions of equipment important to safety described in the LBDs which would be affected by this activity. Failure of the gag to maintain the valve in position would lead to decreased flow to the feedwater pumps. This is an analyzed condition.

Failure of this gag would not create any other type of accident.

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Attachment i to TXX 98021 Page 84 of 110 Evaluation Number SE-97-052 Revision 0 Unit: 1X2 Activity

Title:

Modify DG Starting Air System Configuration and Controls to Establish the safety class interface per ANSI N18.2 & FSAR S3.2 (DM 95-94,95-95 Description of Changes:

The proposed activity (DM 95-094 for CPSES Unit 2, DM 95-095 for CPSES Unit 1, Phase 11) applies to DG Starting Al: system in both units. The subject modifications are planned to be implemented in both units within the short time frame based on the plant availability. Hence, only this evaluation has been prepared to address both units. These modifications include corrective action for ONE Form 95 333.

Modify DG Starting Air system to establish the safety class interface between the SAME Code Section lil, Class 3 DG air receivers and the NNS DG air compressors. Replace (a) compressor unloading NNS solenoid operated valves (SOVs) with ASME Code Section Ill, Class 3/ Class 1E SOVs interlocked to close upon loss of power and/or decreasing receiver pressure (b) associated transformers and fuses in the solenoid electric power supply (c) compressor unloading manual bypass valves (d) booster / delay valve located on the compressor with a time delay relay in the electrical circuit to eliminate breaker trips and other problems leading to unreliable system operation experienced in the past. Add a Class 1E pressure indicating switch source connected to each DG air receiver via a common root valve with the pressure switch and interlocked with the solenoid to ensure a controlled Isolation of the receiver in the event of a line break downstream of the SOVs and/or the compressor. The SOVs and the manual bypass valves will be located away from the compressor to eliminate the vibration problems affecting the valve functional reliability. Piping stress analysis will be performed for the as built configuration to er E e ciructuralintegrity of the affected portion of the system.

Summary of Evaluation:

The safety class interface will be in accordance with ANSI N18.2 and FSAR Section 3.2.

Replacement and relocation of the applicable components and the interlocks will result in functional reliability of the DG starting air system, and isolation of ASME Code Section Ill, Class 3 air receiver from the NNS air compressor under appropriate conditions to ensure at least one emergency DG start. No design basis and LBD requirements and parameters are adversely impacted. No credible potential failure mode will be introduced by this activity.

There is no adverse effect on accidents and malfunctions evaluated in the licencing basis documents and there is no potential for creation of a new type of unanalyzed event due to this activity.

l to TXX 98021 Page 85 of 110 Evaluation Number SE.97-053 Revision 0 Unit: NN2 Activity

Title:

Evaluation of the Effects of the Axial Offset Anomaly on Unit 2 Cycle 3 Operation Description of Changes:

The Axial Offset (AO) Anomaly has been observed during operation of the CPSES Unit 2 Cycle 3 core. The proposed activity was the operation of Unit 2 Cycle 3 with the preence of the AO anomaly. The effects of the AO anomaly on the safety analyses supporting ti, Unit 2 Cycle 3 core design were evaluated herein.

Summary of Evaluation:

The presence of the AO anomaly primarily affects the accident analyses through two parameters: the axial power distribution or shape, and the shutdown margin. Through its presence, the AO anomaly can skew the axial power shape significantly more negative than

- design at the middle-of-cycle and significantly more positive than design at end-of-cycle through its presence, and through its abrupt absence (e.g., crud burst), the AO anomaly can skew the axial power shape significant more positive, in addition, following a plant trip or significant power reduction, the shutdown margin may be eroded due to differences in the neutron flux redistribution effects and due to the loss of concentrated boron (negative reactivity) from the top of some of the fuel assemblies.

The safety analyses supporting the reference Unit 2 Cycle 3 core design include allowances for adverse axial power distributions. These allowances are not allocated to the effects of any specific mechanism. For many of the accident analyses, the AO differences resulting from the AO anomaly are within the allowances for adverse power shapes previously provided in the reference safety analyses. For the remaining analyses, sufficient conservatism in the analysis is provided to offset any adverse affects of the anomaly. Furthermore, the AO anomaly does not affect the integrity of the fuel clad.

This conclusion was dependent on periodic monitoring of the axial offset to ensure that the effects of the AO anomaly have been adequately represented. Subject to this constraint, all applicable event acceptance criteria were demonstrated to remain satisfied; therefore, there was no change to the consequences of the accident as previously ana!yzed nor was there any reduction in the margin of safety as defined by the Technical Specifications.

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.- to TXX-98021 Page 86 of 110 Evaluation Nuraber SE-97-054 i

Revision 0 Unit: NN2 Activity

Title:

Modification of Containment Penetration 2 Mll-0009 for Tubing and Cables for the ' Unit 2 third refueling outage Description of Changes:

Blind flanges from two of three spare pipes in penetration 2 Mil-0009 were removed and local leak rate testing (LLRT) tubing and telecommunication cables were passed through while in MODE 5. Prior to Mode 6 applicability of Tech Spec 3.9.4, the pipes were sected with a suitable penetration seal wiilch ensured the containment atmosphere was isolated from the outside atmosphere. The tubing was isolated in accordance with TS 3.9.4 during core -

alteration and fuel movement inside containment.

Summary of Evaluation:

There are no credible failures introduced by this temporary modification to the containment penetration as described in the updated Final Safety Analysis Report (FSAR). Technical 4

Specification 3.9.4 is to be satisfied while this mod is in place. There is no reduct on in the margin of safety because the containment closure function is to be maintained during core alterations and movement of irradiated fuel during MODE 6 operations.

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Attachment i to TXX-98021 Page 87 of 110 Evaluation Number SE-97-056 Revision 0 Unit: 1N2 Activity

Title:

The Boron Concentration Measurement System is abandoned in place. Descriptions, figures, and table references are deleted from the FSAR.

Description of Changes:

The Boron Concentration Monitoring System (BCMS)is a Non Safety related system which provides continuous inonitoring of the reactor coolant boron concentration. The system is designed for use as an advisory system to provide information as to when additional check analyses are warranted rather than to provide a basis for fundamental operating decisions.

Because this system is not relied on by Operations but does require continual maintenance and calibration, as well as the system contains a neutron source, it was determined to be beneficial to abandon the system in place.

Summary of Evaluation:

The BCMS is a Non-Safety Related system which is not taken credit for in any accident analysis. The BCMS is not part of a control element or control system, nor is it designed for this use. The system is designed for use as an advisory system that provides information as to when additional check analyses are warranted rather than to provide a basis for fundamental operating decisions. Local samples from upstream and downstream of the CVCS mixed-bed and cation-bed demineralizers, as well as downstream of the BTRS demineralizers are able to provide verification of correct CVCS boron concentration. Control of the Boron Thermal Regeneration System can also be baaed on other plant Indications such as control rod position and reactor temperature controller error. There are no safety systems and system important to safety considered that are potentially affected by the implementation cf this activity.

. to TXXc98021 Page 88 of 110 Evaluation Number SE-97-057

~ Revision 0 Unit: 1X2 Activity

Title:

New common compressor for the Service Air System Description of Changes:

This modification provides a replacement common air compressor for the existing two and out-of service compressors for the Service Air System.

Existing line 3-CA-X-003-152G will be revised to accommodate 2 separate service ports, one for active service from new compressor CPX-CACACO-01, and the other for dedicated backup usage. The new compressor will be located near the revised line on an existing concrete foundation west of the Turbine Bldg. The new compressor package willinclude a separate and adjacent oll/ water separator, f

Breaker XB12-1/09/BKR in XB12-1 will be replaced with a 450AT breaker. A new transformer / panel assembly, CPX-EPDPNB-50, will be installed adjacent to the existing disconnect switch CPX-CADSNB-01. A convenience receptacle will be installed adjacent to

' panel CPX EPDPNB-50. New cables and conduits will be installed to provide power to the new compressor and separator.

Service Air compressors, CP1-CACACO-01 and CP2-CACACO-01, are inoperative and will be abandoned in place and removed at a later date. This activity willisolate both compressors with check valves 1CA-0050 and 2CA-0050.

Summary of Evaluation:

The existing Service Air compressors a:e worn past their economic duty life and are currently out of service. The temporary source of service air has been portable air compressors.

Installation of a permanent air compressor will provide the plant a reliable source of air for the Service Air system. The separate service port will allow maintenance to have a dedicated port to use independent of the new compressor.

The locked closed isolation valves at the existing inoperative air compressors will help prevent inadvertent loss of air from the new compressor.

All changes in this modification are non-safety and are not required for safe shutdown. There is no impact on any Technical Specification and therefore no potentialimpact on any margin of safety.

___.__ to TXX-98021 Page 89 of 110 Evaluation Number SE 97-059 Revision 0 Unit: 1N2 Activity

Title:

LDCR SA-97-085 DBD ME-253 R5 & DBD ME-260 R4: Revise RHR Suction Relief Valve Restrictions for Water Solid Operation Description of Changes:

Revise RHR Suction Relief Valve Restrictions for Water Solid Operation to require that at least one residual heat removalline from the reactor coolant loops be unisolated unless the charging pumps are stopped to assure there is a relief path from the RCS to a RHR suction line relief valve. Previously, both were required by the FSAR wording; however, this was and is still inconsistent with the design bases for the cold overpressure mitigation system. The change is a correction and is consistent with Technical Specifications and the current CPSES Precautions, Limitations and Sutpoints (PL&S) which require only one RHR suction relief valve aligned which is sufficient for the LTOPs function.

Summary of Evaluation:

The change to the FSAR is a correction of an error made in response to a question about the low temperature overpressure protection system. The correct wording in accordance with the PL&S and WCAP-10529 requires only one RHR suction line from the RCS to assure there is a relief valvo protecting the RCS when it is water solid. The change has no effect an accident initiation, probability or consequences. The change is consistent with the BASES for the current CPSES Technical Specifications and there is no reduction in the margin 6 safetr d

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_ _ _ _.. _. _ _ _. _ _ _..; to TXX 98021 Page 90 of 110 Evaluation Number SE 97 061 l

Revision 0 4

Unit: NN2 r

j Activity

Title:

Evaluation of the Effects of Power Dependent Axlal Flux Difference Target Bands on Unit 2, l

Cycle 3 End of Life Operation i

i 1

Description of Changes.

]

Appendix 4B of the CPSES FSAR and the Core Operating Limits Report provides the current l

values of the AFD target band as (+5,12)% axial flux difference (AFD) about the target. In 2

order to provide additional flexibility in the operation of Unit 2 Cycle 3 at end of life,it was i

i proposed to implement power dependent AFD target bends that extend linearly from (+5,-12)%

AFD about the target at 100% Rated Thermal Power (RTP) to (+20,17)% AFD about the target l

j at 50% RTP. Between 50% RTP and 15% RTP, the AFD target band remains constant at (+20, t

I 17)% AFD.

Summary of Evaluation The primary effect of the proposed activity was the allowance of more diverse axlal power distributions at power levels below 100% RTP. A second order effect was that the full power axlal power distributions may be silghtly more limiting, relative to the DNBR event acceptance criterion, even though the full power AFD target bands are unchanged. In addition, those reactor physics parameters used in the accident analyses and calculated with a top skewed axlal oower distribution at reduced powers could be affected. However, analyses have been perft <med, using NRC approved methodologies, which demonstrate that sufficient margin is avallat,le in the calculations such that the reactor physics parameters esed in the accident 4

analyses remain valid. Sensitivity studies have shown that the combination of full power axial i

powar distributions,100% RTP, and full power F-delta H remains more limiting, relative to the DNBR event acceptance criterion, than any consistent combination of part power axial power distribution, part power, and part power F delta H. The fuel assembly mechanical design analyses were also shown to remain valid. Because all relevant event acceptance criteria have been demonstrated to remain satisfied, there is no ch6nge to the consequences of the accident as previously analyzed nor is there any reduction in the margin of safety as defined by the plant i

i Technical Specifications.

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- - - -..... - - to TXX 98021 i

Page 91 of 110 Evaluation Number SE. 97 065 i

Revision 0 i

Ur.it: 1N2 Activity

Title:

i Evaluation of the Use of an Average Tavg Function in the CPSES NSSS Control Systems l

Description of Changes:

Ihis activity is the replacement of the auctioneered high Tavg unit with an average Tavg unit, i

ThS output signal of the auctioneering/ averaging unit provides an input to the Reactor Coolant System, the Pressurizer Water level Control System, and the Steam Dump System. This j

activity is expected to have neg!!gible effects on the operation of the Pressurizer Water Level l

Control System and the Steam Dump Control System, result in reduced stepping of the control rods when in the Reactor Control System is in the automatic mode of operation, and allow the plant to be operated at a slightly higher temperature, conalstent with the safety analyses, i

Summary of Evaluation:.

For CPSES Unit 2, which has no significant loop to loop asymmetries, the replacement of the auctioneering unit with an averaging unit is relatively transparent during normal operation. For CPSES Unit 1, the effects of the upper plenum flow anomaly cause continuous control rod aosilla changes, when in the automatic mode of operation, due to the step changes in the one cop's (and auc tioneered high) Tavg Induced by the present and subsequent absence of the upper plunum flow anomaly. Because another loop Tavg is complementary to the affected loop Tavg, the average Tavg is relatively o 'affected by the upper plenum flow anomaly.

3 The use of an average Tavg signal allows the bulk RCS temperature to be closer to the value assumed in the safety analysis. Even though the resulting temperature is higher than If the

~

j auctioneering unit is used, the value remains within the assumptions of the accident analyses.

All relevant event acceptance criteria remain sat!sfied.- The effects of failures in single Tavg loops are different and result in different magnitudes of ensuing transients; however, all failure modes are within the bounds of the original failure analysis and within the bounds of the accident analyses. Because none of the safety analyses are adversely affected by the

' proposed activity, and the proposed activity does not result in the reduction of any failure 4

points, it is concluded that the margin of safety is unaffe rted by the proposed activity.

' t)TXX 98021 Page 92 of 110 Evaluation Number SE 97 066 Revision 0 Unit: IN2 l

Activity

Title:

REVISc TABLE 5.2 5 LITHlUM BORON RANGES i

Description of Changes:

Revise Table 5.2 5, changing the lithium band from ppm lithium to reference a plant pH(t) range, revise the Boric Acid (ppm Boron) range reflecting the boron required for operating with higher energy cores, and remove references to electrical conductivity, solution pH, and Total Suspended Solids.

Summary of Evaluation:

Revising Table 5.2.5 does not result in a previously unanalyzed accident or increase 63 probability of an analyzed accident. The impact on plant materials, systems, components, and fuel cladding is expected to be negliglHe.

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to TXX 98021 Page 93 of 110 Evaluation Number SE. 97 067 Revision 1 Unit: 1N2 Activity

Title:

RWST !_ow Low Setpoint Change to Enable Containment Spray Switchover Prlor to Empty Alarm Description of Changes:

The design modifications including RWST Low Low setpoint and valve motor operator gear changes and procedure changes comprise corrective action for LER 97 002. The setpoint and gear changes a e being made to enable the switchover of ECCS and containment spray without stopping the pumps during transfer from the RSWT to the emergency sumps The supporting calculations and Final Safety Analysis Report (FSAR) changes ara required to show the design meets the design bases for containment spray and ECCS transfer. The procedure changes are required to implement the revised analyses. This is consistent with the design basis for the ECCS and containment spray system and the accident analysis. However, the minimum RWST volumes for injection into containment are affected. Containment analyses, including pressure / temperature, pH, NPSH, and hydrogen generation were updated and the impact on equipment qualification evaluated. The time to switchover to hot leg recirculation was also updated to assure long term cooling. This activity may be implemented under 10CFR50.59 because the change is conservative with respect to the TS 3/4.3.2 setpoint and compliance will be maintained, in addition, this change is corrective action under 10CFR50, Appendix B and the schedule for implementation falls under prompt corrective action requirements.

Summary of Evaluation:

The RWST level setpoint calculation and all affected calculations and setpoints were revised to demorist"to the ability to complete containment spray pump switchover without stopping the

pumps, ne Iro 'act on LOCA and MSLB accidents were evaluated. The effect on equipment qualification, hydrogen generation and long term NPSH for ECCS and containment spray recirculation from the containment eme'

' r sumps were determined to be acceptable. The emergency operatin; procedures are be

. elsed to reflect the new analyses and setpoints.

There are no credible failure modes created by implementation of th:s activity. No hardware is being added, de eted or modified except for the motor operator gear change on the RWST to containment spray isolation valves under the MOV program. The setpoint change does not introduce a failure mode. Equipment qualification impacts were evaluated and documented to be acceptable. Hydrogen generation calculations were updated and found to meet the 3.5%

limit.

The affected acceptance limits for the RNST is to provide sufficient volumes for ECCS and containment spray transfer without stopping pumps and for NPSH for emergency recirculation.

The CPSES SER section 0.5.2 states that the containment spray system should be capable of continuous operation for at least two hours and be capable of switchover without stopping the pumps. The setpoint changes, gear changes and procedule changes ensure the licensing basis is restored.

The design modifications including RWST Low Low setpoint and valve motor operator gear changes and procedure changes complete corrective actions for LER 97-002.

. _ = - -.

_ to TXX 98021 Page 94 of 110 Evaluation Number SE-97 069 Revislen 0 Unit: IN2

-Activity

Title:

Summarize FSAR Table 10.4 4 data into FSAR Section 10.4.6.5 4

)

Description of Changes:

Delete Table 10.4 4 (Sheets 1 thru 11) by summarizing significant data into section 10.4.6.3.

Summary of Evaluation:

There are no identified credible potential failures modes for the affected structure, system, or com)onent, that could be introduced by implementation of the activity. This activity does not invo:ved an Unreviewed Safety Question. This activity will have no effect on accidents and malfunctions evaluated in the LBD nor increase the potential for croation of a new type of unanalyzed event. This activity removes data which is not required by RG 1.70 which states the following for section 10.4.0: "The design basis for the condensate cleanup system should include the fraction on condensate flow to be treated, impurity ievels to be maintained, and design codes to be applied. The evaluation of the condensate cleanup system should include an analysis of demineralized capacity and anticipated impurity levels, an analysis of the contribution of impurity levels from the secondary system to reactor coolant system activity levels, and performance monitoring. Provisions for the control of chloride ion and other contaminates should be described." Section 10.4.6 of the CPSES FSAR restates any pertinent data removed from Table 10.4 4.

f Attachment i to TXX 98021 Page 95 of 110 Evaluation Number SE 07 073 Revision 0 Unit: NN2 Activity

Title:

Installation of a 75.5 Ft. tall pole inside the PA & stringing of 3/8" steelshield wire from pole to dead end tower for U2 Description of Changes:

The DM activilles are as follows:

a) Drilling of a 4' diameter by 25 ft deep hole inside the protected area. The coordinates of the hole are N 10031.0, E 10292.5 & it is located east of the concreto drainage ditch, b) Installation of reinforcing steel and placement of concrete in the above hole for the pole foundation.

c) Assembly and erecting the 75.5 ft galvanized steel pole on the above foundation,

[

d) Stringing of 250 feet (approximately) of 3/8" diameter galvanized steel shield wire from the new pole to the dead end tower for the Unit 2 Main transformers.

e) installation of four,6" diameter bumpers around the pole to preclude incidental damage to the pole.

Erection of the pole will be done at power. However, the stringing of the shield wire will be performed during 2RF03 when the 345 kV sides of the Unit 2 Main Transformers are deenergized.

Summary of Evaluation:

Engineer ng review of the DM design has determined that there are no Interferences above grade or under ground at the pole location. This review has also determined that in the unlikely event of the pole falling due to severe weather conditions, the pole itself will not come in contact with any of the plant structures, systems, components or overhead conductors. In this unlikely event, the shield wire could short the 345 kV busses on the output side of the Main Transformers (2MT1 and 2MT2). The 345 kV circuit breakers in the switchyard would trip to orotect the faulted circuit. Shorting of the 345 kV busses for the Main Transformers would however, cause the loss of Unit 2. Tripping of the Unit would not impact continued availability of offsite power to 6.9 kV Class 1E Busses for any of the Units. Therefore, the shorting of the 345 kV lines would cause Unit 2 to shut own but it will have no adverse effect on the safe shutdown of Unit 2 and the safe operation of Unit 1.

Based on the above evaluation,it is concluded that implementation of DM 97-042 does not result in an unraviewed safety question.

_ to TXX 98021 Page 96 of 110 Evaluation Number SE.97 074 Revision 0 Unit: 1X2 Activity

Title:

Procedure changes to sparge the Boric Acid Tanks with nitrogen to reduce the concentration of oxygen in solution l

Description of Changes:

Sparge the Boric Acid Tanks (BATS), CPX-CSATBA-01 and CPX CSATBA-02, with nitrogen to reduce the concentration of oxygen in solution. The point of entry to the BATS is through a tank level transmitter sene'ng line drain.

Summary of Evaluation.

1 From a Chemistry standpoint, it is more desirable to have nitrogen than oxygen in the BAT fluid While a BAT is being sparged with nitrogen through a level transmitter sensing line drain, the level transmitter might be considered nonfunctional since it would indicate slightly higher than actual tank level due to the back pressure form the nitrogen entering the tank. There is a redundant level transmitter for continual monitoring of tank level, and the time of the sparge operation is being limited to seven days so the two level transmitters can be compared for continued confidence tnat the indicated level represents actual level on both indicators. The level transmitter being nonfunctional cannot directly result in accidents or equipment malfunctions. The nitrogen will be introduced at a rate of no more than 1 SCFM at a location physically higher and on the circumferenc9 of the tank approximately opposite the boric acid transfer pump suction from the tank, so n" ogen gas will not be entrained in the suction of the pumps. The low flow rate of the nitrogen also will not present a concern for overpressurlzing the tank, which is vented to atmosphere through a 4" diameter vent line. The nitrogen will be admitted through a normally closed drain valve, which is also the point of a piping class break frorr 3 to 5. If a leak were to occur In the non code nitrogen rig such that the BAT started to drala though the break, compensatory actions are in place to prevent the tank from draining below the minimum required TS level.

. to TXX 98021 Page 97 of 110 Evaluation Number SE. 97 076 Revision 0 Unit: 1N2 l

Activity

Title:

BAD STATOR RTD'S FOR 6.6KV MOTORS CAN BE ABANDONED IN PLACE & MAY BE REPLACED WHEN MOTORS ARE REWOUND Description of Changes:

in case a stator winding (c; 6.6 kv motor) temperature detector (RTDs) falls, the RTDs ccn be abandoned in place and may be replaced when the motors are rewound.

Summary of Evaluation:

i Motors are provided with two tempereture detectors (RTDs)In each phase for stator winding temperature monitoring purposes where required. These RTDs are not used for any motor t

protection, it is possible that dudng life of a motor, these RTDs may go bad. Because these RTDs are used only for monitoring purposes capability of motor to perform its function. The RTDs can be abandoned in place and may be replaced when the motors are rewound. Plant i

4 Computer monitoring capability, which is a non 1E function, may not be available if all the RTDs in motor stator winding have failed.

This activity does not affect the probability or consequences of accidents or malfunctions of equipment important to safety already analyzed in the LBDs and does not affect Tech. Specs.

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Attachment i ta TXX 98021 Page 98 of 110 Evaluation Number SE. 97 077 Revision 0 Unit: NN2 Activity

Title:

j Evaluation of the effects of a +5 AFD target on Unit 2 Cycle 3 EOL Operation Description of Changes:

Previous evaluations of the ehacts of the Axial Offset (AO) Anomaly on Unit 2 Cycle 3 operation were considered valid for full power axial flux difference targets of up to +3% delta 1.

Because of the effects of the AO anomaly, the sxial flux difference is drifting positive at the i

end of cycle. The activity is the evaluation of the effects of full power axial flux difference targets of up to +5% delta I on the safety analyses supporting Unit 2 Cycle 3 operation. The evaluation is limited to exposures of 20,358 MWD /MTU and greater.

r Summary of Evaluation:

The primary effect of the proposed activity is the allowance of more top skewed axial power l

distributions. The axial power distributions may be slightly more limiting, relative to the relevant event acceptance criteria, than those previously considertd. To offset the adverse effects associated with these power distributions, the values of the Fq and F delta h peaking

. factors are limited to 2.40 anri 1.53, respectively. Analysis has shown that these limits the peaking factor limits is sufficient to offset the adverse effects of the more top skewed axial power distributions. All analyses were performed using NRC-approved methodologies. T'.e fuel assembly mechanical design analyses were also shown to remain valid. Because all relevant event acceptance criteria have been demonstrated to remain satisfied, there is no change to the consequences of the accident as previously analyzed nor is there any reduction in the margin of safety as defined by the p' ant Technical Specifications.

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to TXX-98021 Page 09 of 110 Evaluation Number SE. 97 078 Revision 0 Unit: 1X2 Activity

Title:

i Change Sotpoint of Product H2 Alarm and Product O2 Limit for Catalytic Hydrogen Recombiners, Gaseous Waste Processing System, LDCR SA 97133 Description of Changes:

1. Allow the change of the high hydrogen product alarm set point to 2% Product hydrogen is the concentration of hydrogen gas being discharged from the reaction chamber of the catalytic hydrogen recombiner, ll. Allow operation of the catalytic hydrogen recombiners with product hydrogen concentration exceeding the current alarm set point of 0.25%, but less than 2% until the high hydrogen product set point change is implemented.

Ill. Allow the change of the high product oxygen setpoint to 2500 ppm. Product oxygen is the concentration of oxygen gas being discharged from the reaction chamber of t e catalytic h

hydrogen recombiner. This change also changes the set point for automatic isolation of the oxygen supply valve to 2500 ppm product oxygen.

Summary of Evaluation:

The only safety function provided by the Gaseous Waste Process System (GWPS)is the retention of radioactive gases for decay. The system is designed so that a single failure of the system will not result in an offsite dose which exceeds the 0.5 REM criterion that is described in the NRC Branch Technical Position ETSB 11-5 and the site boundary requirements of 10CFR100. To meet these requirements the total activity in a single gas decay tank is limited to 200,000 Cl noble gas by TS 3.11.2. A limit of less then or equal to 3% oxygen is imposed whenever hydrogen concentration exceeds 4% to preclude an explosive / flammable gas mixture in the GWPS to provide assurance that releases of radioactive materials will be controlled in conformance with the requirements of GDC 60 of 10CFR50 Appendix A. Automatic control features are provided to maintain the concentration of oxygen below the flammability limit of 3% including isolation of source hydrogen and oxygen. Automatic control features are also provided to maintain the concentration of feed hydrogen to less then 95 The design purpose of the high hydrogen product alarm is to provide indication of possible high hydrogun feed concentration to the catalytic hydrogen recombiner, recombiner reactor malfunction, or loss of oxygen supply. This alarm provides indication only and does not feed circuitry supplied for the isolation of influent hydrogen or oxygen sources to preclude exceeding explosive / flammability gas limits. Operation of the system below the current set point of 0.25%

3roduct hydrogen is impractical due to the presence of helium in the influent gas stream and nstrument accura.

Hydrogen and helium both permeate the detector's gas permeable membrane. Helium is detected as hydrogen by the gas analyzers due to its ability to permeate the membrane, and its presence registers as hydrogen even when all hydrogen has been reacted. The current setpoint of 0.25% for the high product hydrogen alarm is unncessarily low, and normal operation of the system causes this alarm to be locked in continously.

Independent control functions currently limit the product oxygen supply to the GWPS. Raising the high hydrogen product setpoint from 0.25 % to 2% allows the alarm to be functional under normal system operat!on. It provides adequate allowance for the impact of helium in the influent gas stream and the limits of exis'ing system Instrumentation without challenging the design basis of the system.

Influent gas admission to the GWPS orlor to implementation of the high hydrogen product alarm setpoint change is acceptable 'nce technical specifications only require one hydrogen

._ to TXX 98021 Page 100 of 110 analyzer channel for equipment operability.

The design function of the high ox/ gen product setpoint is 'o maintain the concentration of oxygen within the system to less than or equal to 3% in the event that hydrogen concentration exceeded 4%. The basis of this controlls to preclude a gas mixture within the Waste Gas Holdup System which exceeds the flammability limits of oxygen and hydrogen. Raising the high oxygen product set point to 0.25% (2500 ppm) does not challenge the design basis.

Operation of the catalytic hydrogen recombiner with a slightly higher concentration of product oxygen allows the operators to more completely react the hydrogen within the tanks or system and obtain an improved helium compensation factor. Raising the product oxygen setpoint from 60 ppm to 2500 ppm (0.25%) continues to require that the system be operated in an oxygen lean environment and does not conctitute a challenge to the oxygen flammability limit.

Expecting normal system operation to maintain the product oxygen less than 60 ppm (0.006%)

is an unnecessary conservatism compared to Technical Specification requirements.

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, to TXX 98021 Page 101 of 110 Evaluation Number SE 97 079 t

Revision 0 Unit: 1N2

/

Activity

Title:

Remove FSAR Administrative limits on heatup & cooldown rates for the Reactor Coolant

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4 System and pressurizer Description of Changes:

Remove administrative limits on heatup and cooldown rates of the reactor coolant system (RCS) and Pressurizer (PZR), respectively. The changes facilitate shutdown and startup from -

l outages by allowing full use of the less stringent limitations imposed by the Technical

- Specifications.

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Summary of Evaluation

• i The Technical Specifications limit the RCS heatup and cooldown rates to 100F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 3eriod, the PZR heatup rate to 100F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, and the PZR cooldown rate to 200F n any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.- The SAR describes administrative limits for these parameters as RCS heatup and cooldown rates of 60F/hr and PZR heatup and cooldown rates of 100F/hr. Removal

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of the SAR administrative limit on heatup and cooldown rates of the RCS eliminates an unnecessary burden on plant operation that is not founded Iri analysis nor required by regulation. In actual 3ractice, the limit of RCS temperature change rate of 100F per hour imposed by Technica Specifications will not be attained because of other physical and operational plant limitations. Removal of the SAR administrative limit on heatup and cooldown rates of the PZR is of no consequence since the basis for the limit was only that it was an assumption in an unrelated cooldown analysis, for which had the temperature change rate of

- 200F per hour imposed by Technical Specification been used, it would have provided additional margin. Operating per the Technical Specification Imposed limits is consistent with the design of the RCS and PZR.

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-. _ - -. - - - to TXX.98021 Page 102 of 110 Evaluation Number SE-97-080 Revision 0 Unit: IN2 1

Activity

Title:

REPLACE RVLIS PROBE Description of Changes:

The Unit 1 Train B Reactor Vessel Indicating System (RVLIS) Probe is being replaced by DM J

95 065 because the current probe has sensors 18,38,4B, and 7B out of service. Thh meets the Tech Spec limits for operability, however, if one more sensor is lost, this train will be considered inoperable. This probe is a direct replacement to the original probe except that a 4

single mutlpin connector to a single head lift rig cable replaces eight connectors to eight head lift rig cables. Allowing the use of Stainless Steel clad mineralInsulated RVLIS cable is acceptable for R.G.1.75 separation requirements which will allow the cables to be installed without fire barrier material (sittemp) which wl!I reduce manhours, man rem exposure and outage time during every refueling outage.

Summary of Evaluation:

The RVLIS Probe and head lift rig cables are Class 1E components in the RVLIS system which arovides the Control Room indication to measure the reactor coolant inventory in the upper lead and plenum regions of the reactor vessel. RVLIS is considered Accident Monitoring Instrumentation and complies with RG 1.97. However, since the new probe is a direct replacement for the old probe and the new head lift rig cable is equivalent to the eight original head lift rig cables with an additional stainless steel flexible conduit around them the qualification of functionality of the RVLIS system will not be affected. Metal Clad cable is described in the FSAR but not the type used for the existing or the new mineralinsulated cable and the stainless steel flexible conduit and the inner stainless steel cladding of the new RVLIS mineralinsulated cable and the stainless steel cladding of the Unit 2 and Unit 1 Train A RVLIS mineralinsulated cables are each equivalent to the stainless steel flexible conduit for separation purposes as referenr.*A in the FSAR. The addition of the stainless steel clad mineralinsulated RVLIS cables a ;he FSAR as equivalent to an enclosed raceway will not adversely impact the operation of the units. There are no safety systems and systems important to safety considered that are potentially affected by the implementation of this activity.

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- to TXX 98021 Page 103 of 110 Evaluation Number SE. 97 081 Revision 0 Unit: NN2 Activity

Title:

Unit 2 Two Train Componen'. Cooling Water (CCW) Outage During Unit 2 Third Refueling Outage (2RF03)

Description of Changes:

A two train CCW outage in Unit 2 during 2RF03 is required to perform maintenance work activities on the CCW system. During the time period that Unit 2 CCW is not available, CCW cooling to required Unit 2 equipment, specifically the Spent Fuel Pool Coollng and Cleanup-System (SF) heat exchangers, will be provided from Unit 1 CCW. This Safety Evaluation addresses the plant conditions during the Unit 2 two train CCW outage evolution, what responses are required for various possible accident or failure scenarios, and the Defense in-Depth provisions that will be in place to perform this activity. This Safety Evaluation is required for the potential deviations from the plant processes and configuration as described in the FSAR for Unit 1 and SF.

Summary of Evaluation:

This Safety Evaluation evaluates the plant condition required to perform maintenance on a CCW valve that is a single point of Interface between the two Unit 2 trains. This requires that both Unit 2 CCW trains be shutdown. CCW cooling to required Unit 2 equipment will be provided by Unit 1 CCW. This evaluation considers the following potential plant accidents or failures; loss of SF cooling due to SF equipment failure; loss of SF cooling due to CCW equipment failure; LOCA in Unit i resulting in of loss of SF; LOCA in Unit 1 coincl dent with a Single Active Failure; Station Blackout; Fire causing loss of SF cooling; CCW pump run out under various plant configurations. This evaluation shows that during the time period that Unit 2 CCW la not available, with a LOCA in Unit 1 coincident with a Single Active Failure in Unit 1, steps can be taken to cool the SF pools to within their abnormal maximum design condition of less inen 212 degrees F. This can be accomplished with the Defense-in Depth provisions that will be implemented as part of the 2RF03 outage plan. The activity does not involve an Unreviewed Safety Question.

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to TXX 98021 Page 104 of 110 Evaluation Number SE-97-082 Pevision 0 Unit: 1N2 Activity

Title:

FSAR corrections for Hot Shutdown Panel, Shutdown Transfer Panel, ar.d Control Room Layout.

Description of Changes:

These changes provide corrections and clarifications to the following FSAR descriptions: 1)

Section 7.4.1.3.1 Paragraph 7 is clarified to reflect that the Hot Shutdown Panel's (HSP) cabinet doors are locked and alarmed in the Control Room while the surrounding enclosure door is locked but not alarmed; 2) Table 7.4-2 is corrected to delete Indication Fl 1776 (Bor!c Acid Filter Recovery Line - Flow) and add indication F1-183B (Emergency Borate Flow); 3)

Table 7.4 2 typographical errors are corrected by changing A-1EG2-1L to A-1EA21L and HC-61B to HC 618; 4) Table 7.4-3 is corrected to delete switch HS-XT (Water chiller control remote local) and add HS 6710B (Safety Chill Water Chiller Bypass Lockout); 5) Figure 7.7-14A is revise.o add the Leading Edge Flow Meter in place of removed RM 21 Printer and delete removed RMS equipment.

Summary of Evaluation:

Change 1) to Section 7.4.1.3.1 Paragraph 7 clarifies the FSAR description consistent with the as built plant hardware configuration. The cabinet doors of the Hot Shutdown Panel's (HSP) are locked and alarm in the Control Room. The surrounding enclosure of the HSP is locked but does not alarm in the Control Room. The purpose of the HSP surrounding locked enclosure is for access control to prevent inadvertent plant equipment manipulation from the HSP. Access to the HSP is administratively controlled by Operations personnellocated in the Control Room.

Therefore, an alarm is not needed on the HSP surrounding enclosure to accomplish this function and the absence of this alarm has no impact on plant safety. This change has no impact on Safe Shutdown. Change 2) corrects Table 7.4 2 by deleting F11776 (Boric Acid Filter Recovery Line - Flow) which is not a valid plant parameter based on review of the Master Equipment List, Emergency Response Guidelines (ERGS), and plant drawings M1-0257, M1-0255, and M2-0255. Fl-1838 (Emergency Borate Flow)is added to Table 7.4 2 because the Boric Acid Filters are in series with indications 1-Fl 1838 and 2 FI-183B during Emergency Borate operations and therefore would have the same flow. Emergency Borate Flow is a critical plant parameter which is monitored from the HSP. Change 3) corrects typographical errors which are considered to be trivial changes in accordance with CPSES guidelines.

Change 4) deletes switch HS-XT (Water chiller control remote local) from Table 7.4-3 and adds HS-6710B (Safety Chill Water Chiller Bypass Lockout). This change corrects Table 7.4-3 to reflect previously approved Engineering Change Notice ECN-208 made to comply with Appendix R requirements described in DEP37 Rev.1. This change does not remove the ability to remotely operate the Safety Chill Water Chiller but diers the circuitry to comply with Appendix R. This change has no impact on Safe Shutdown. Change 5) updates Figure 7.7-14A consistent with changes to the control room lay out. Changes to the Radiation Monitoring System and Meteorological System were previously evaluated by Safety Evaluation SE-91-082.

The Leading Edge Flow Measuring System Cabinet is a non-safety related system previously described in FSAR Teble 17A-1. since Change 5) has been previously evaluated, it will not be evaluated in the evaluation.

hfachment 1 to TXX 98021 Ps49105 of 110 Evaluation Number SE 97 083 Revision 0 Unit: NN2 Activi.ty

Title:

STEAM GENERATOR 1 SHELL PENETRATION Description of Changes:

A foreign object was found on the hot leg side on top of support plate "L"in Steam Generator 1 of Unit 2. The object is located on the periphery of the tube bundle next to tubes R49C53 and R49C54. The object wore into these two tubes, one to ap 3roximately 70% depth. A video probe inspection of the part was performed, however, retr eval of the object was not considered feasible from the existing 2.5 inch Access Port #1, which is located approximately 90 degrees away from the foreign object. An inspection port nominally 2.5 Inches in diameter will be installed in steam generator 1 to facilitate removal of this object.

Summary of Evaluation:

The steam generator shell penetration and closure hardware are designed, analyzed and constructed in accordance with the ASME Code. Stresses and fatigue usage in the shell penetration and in the closure hardware are within the AMSE Code Allowable values. The

)ressure boundary of the steam generators is maintained. The modification does not change nspection requirements, nor does it prevent inspections currently required. Design features and procedural precautions minimize the possibility of loose parts. The integrity and 4

performance of the steam generators are unaffected. No other safety related equipment is d

affeeted.

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Attachmont 1 to TXXo98021 Page 103 of 110 i

Evaluation Number SE 97 085 Revision 0 l

Unit: NN2 l

Activity

Title:

Temporary New Fuel Storage Container Description of Changes:

A temporary storage container will be installed in the wet cask pit in the Fuel Building to store a damaged new fuel assembly. The container will be non safety related and non selsmically designed. The internal dimensions will be larger than normal for fuel assembly storage cells.

The location is isolable from the spent fuel pools by two refueling gates. Once the assembly is in the container, the cask pit may be drained and the container may be moved to another area suc,h as the decontamination area new fuelinspection area of the fuel building.

Summary of Evaluation:

There is no impact on spent fuel storage due to the location of the temporary container. The only potentialis for further damage tc, the new fuel assembly. Evaluatior's have been performed that conclude that under worst case condittoris Keff would not exceed 0.95. The container has a spacer at the top which ensures another new fuel assembly cannot be closer than 12 inches to the container preserving the 21 inch center to center separation requirement.

The temporary container will be austenitic stainless steel. Therefore, the use of this temporary container does not involve an unreviewed safety question.

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Evaluation Number SE.97 086 j

Revision 0 l

Unit: IN2 i

Activi

Title:

I REVIS PRIMARY AND BACKUP F'ROTECTION DETAILS AND TO PROVIDE LIMITS AND RESTRICTIONS FOR CRD MOTOR GENERATOR SETS NEUTRAL GROUND SW i

)

Description of Changes:

1. Provide limitations and restrictions for CRD motor generator sets neutral grounding switch i

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to be kept open when the motor generator sets are feeding power to the control rod drive system.

2. Revise penetration protection details as follows:

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a) Take credit for fuses as primary protection device in control rod lift coil circuits. b) Delete 3rimary and backup protection details for solid state Isolation cabinet and containment water i

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.evel transmitter circuits. c) Delete containment door limit switches penetration description.

i Summary of Evaluation:

i The ground fault current in the CRD coils power circuits limited to 3.51 amps by the 50 ohm 1

resistor in the neutral grounding circuit of the motor generator sets. CRD coils power circuit penetrations are self protected at this value of the ground fault current. The closing of the motor generator sets neutral grounding switch bypasses the grounding resistor and in this mode the ground fault will be limited by the current limiting features of the CRD coil circuits.

The current limiting features of the coil circuits limit the current to a value less than the

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continuous rating of the penetration conductors. The closing of the switch concurrent with a sitting ground fault in a coil circuit upstream of negative leg fuse may leave the penetration conductor unprotected if a single failure of the positive leg fuse is postulated and the non Class i

1E current limiting feature of the circuit is assumed to fall. Therefore, the neutral groundmg switch of the CRD motor generator sets shall be kept open when the motor generator sets are feeding power to the CRD system. This configuration will assure that the CRD coils power i

circuits penetration conductors shall remain self protected from any ground fault.

i 50 amps fuses, one in each positive and negative leg of CRD lift coil circuits, provide adequate protection from a line to line fault for #4 AWG senetration conductors used in these circuits.

Therefore, credit is taken for the 50 amp fuse :n each leg of the circuit to provide primary and backup protection for #4 AWG penetration conductors.

i Fault current values in Solid State Isolation Cabinet and Containment Water Level Transmitter circuits are less than the continuous rating of the penetration conductors, as such these circuits do not require primary and backup penetration.

Penetration protection details for Containment Door Limit Switch circuit are deleted because the circuit has been spared and the cables at penetration conductors have been determined.

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_ __ _. _ _. _ _ _ _ _ _. to TXX 98021 I

Page 108 of 110 Evaluation Number SE.97-088 Revision O Unit: 1N2 Activity

Title:

Operate the Positive Displacement Charging Pump with the suction stabilizer isolation valves in the normally closed position i

Description of Changes:

Through testing performed by Westinghouse (Reference WPT-15802),it has been determined that the flow capacity of the positive displacement charging pump (PDP) suction stabilizer gas vent line is very low. Even leakage past the seat of suction stabilizer gas supply control valve, u 8204, may be greater than the capacity of the vent line. This creates the potential when the i

PDP is operating to fill the suction stabilizer and eventually the common centrifugal charging aump (CCP) suction header with gas. This could cause the PDP and both CCPs to be noperable, in order to prevent this from occurring, it is proposed to revise the method of operation for the gas supply valves, u 8210A and u 8210B. The valves will be maintained in a normally closed position at all times, including when the PDP is in operation. When the stabilizer reaches high level, as indicated by the red light on gas control valve, u 8204, control switch, the gas supply valves will then be manually opened to lower water level below the high level. Once levelis restored to the normal range, the gas supply valves will again be returned to the closed position. If system hardware is functioning properly,it is expected thut the need to lower the PDP suction stabilizer level will be on an infrequent basis.

Summary of Evaluation:

i There is no credible potential failure mode created by this activity. With the PDP suction stabilizer gas supply check valve and two in series closed gas supply valves preventing leakage of gas out of the suction stabilizer, the only credible way the stabilizer can go solid is by the very slow process of gas going into solution. There is thus adequate time for operator action to open the gas supply valves manually, rather than relying on automatic action to supply gas to the suction stabilizer. Site procedures ALM 0061 A/B already require that the PDP be secured if the HI HI alarm is received, thus removing the source of energy that could damage system piping if the suction stabilizer did go solid. The design basis for the level set point at which the gas stpply control valve, u-8204, gets an open signalis arbitrary, the only requirement being that the gas supply valve should open before the HI HI level alarm set point is reached, so operator manual opening of the gas supply valves after u 8204 is open is consistent with the design basis. This activity does not involve an Unreviewed Safety Question or require an amendment to the Technical Specifications. It does not adversely affect any system used for accident mitigation, will not impact plant response to system failure, and will meet system design requirements.

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-- - - -_ _-. --- to TXX 98021 Page 109 of 110 Evaluation Number SE-97-090 Revision 0 Unit: 1N2 Activity

Title:

Delete ref to the FSAR statement that the containment pressure relief line is located in en f

enclosed room which does not contain SR equipment i

Description of Changes:

This change deletes reference to the FSAR statement that the containment pressure relief line is located in an enclosed room which does not contain safety related equipment, Summary of Evaluation:

The evaluation Indicates that a separate enclosed room is not required to protect nearby safety related equipment from affects of escaping air and steam following a postulated LOCA 4

because: 1) the 18 nch pressure relief line is orificed to 3 inches; 2) a control room alarm i

alerts the operator if pressure exceeds normal maximum containment operating pressure,3) both pressure relief Containment i alation valves automatically close within three seconds after receiving the signalif high radiation is detected by the containment particulate lodine-gaseous monitor, on a phase A isolation signal, or on containment ventilation isolation 4) leak rate testing of the containment penetration and the penetration isolation valve response time I

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Attachment i to TXXe98021 Page 110 of 110 Evaluation Number SE. 98 003 Revision 0 Unit: 1X2 Activity

Title:

Revise FSAR Description to delete Indication only alarms of 6% feed H2 co..c. and 15 ppm product O2 conc. from H2 recombiner, GWPS Description of Changes' A 1993 plant design modification replaced the Gaseous Waste Processing System (GWPS) oxygen and hydrogen analyzers and various controls on the system. However, the new analyzers did not have the capability to supply all of the previous alarms and alarm / control points. As a result, two (2) indication only alarms on the GWPS hydrogen recombiner (intet HI hyrogen concentration at 6% by volume and discharge HI oxygen concentration at 15 ppm by volume) were removed from the system; however, the FSAR (Section 11.3) description was not updated. This activity revises the FSAR to reflect the as built configuration of the C6talytic Hydrogen Recombiner of the GWPS and evaluates the removal of the subject indication only alarms from the system.

Summary of Evaluation:

The safety function provided by the GWPS is the retention of radioactive gases for decay.

i Criterlon in the NRC Branch Technica! Position ETSB 115 and the site boundary requirements of 10 CFR 100 require that the GWPS be designed so that a single failure of the system will not i

result in an offsite dose in excess of 0.5 Rem. To meet these requirements the total activity in a single gas decay tank is limited to 200,000 Cl noble gas by Technical Specification 3.11.2. A technical specification limit of less than or equal to 3% oxygen is imposed whenever hydrogen exceeds 4%. This limit precludes an ex alosive/ flammable gas mixture in the GWPS and provides assurance that releases of rad oactive materials will be 2ntrolled in conformance with the requirements of GDC 60 of 10CFR50, Appendix A. Automatic control systems are provided in the GWPS to maintain the concentration of oxygen below 3% and hydrogen to less that 9%.

The alarms eliminated in this activity provided preliminary indication only that the system is operating at or slightly beyond its limits of 100% recombination capacity. The subject alarms also provided preliminary indication of high hydrogen feed concentration to the catalytic hydrogen recombiner, reactor malfunction, loss of the preheater, or a loss of oxygen supply and continued operation may result in the automatic controls shutting the system down.

The alarms that were eliminated provided ind, cation only and did not have circuitry supplied for control of the preheater or the Isolation of influent hydrogen or oxygen sources to preclude exceeding explosive / flammability gas limits. The alarms and control points retained in the GWPS provide the capability to continue to meet requirements as noted above.

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