ML18026A258

Forwards Summary Rept of Safety Evaluations Approved During Period from 920101-921231 for SSES Unit 1 & 2

ML18026A258
Person / Time
Site:	Susquehanna
Issue date:	12/22/1993
From:	Byram R PENNSYLVANIA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
PLA-4061, NUDOCS 9312300272
Download: ML18026A258 (130)

Text

~';ACCELERATED D'RIBUTION DEMONST+.TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9312300272 DOC.DATE: 93/12/22 NOTARIZED: NO DOCKET FACIL:50-387 Susquehanna Steam Electric Station, Unit 1, Pennsylva 05000387 50-388 Susquehanna Steam Electric Station, Unit 2, Pennsylva 05000388 AUTH. NAME AUTHOR AFFILIATION BYRAM,R.G. Pennsylvania Power 6 Light Co.

RECIP.NAME RECIPIENT AFFILIATION V Document Control Branch (Document Control Desk)

SUBJECT:

Forwards summary rept of safety evaluations approved during period from 920101-921231 for SSES unit 1 & 2.

DISTRIBUTION CODE: IEOID COPIES RECEIVED:LTRIL ENCL I SIZE:

TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-2 PD 1 1 CLARK,R 1 1 D INTERNAL: ACRS 2 2 AEOD/DEIB 1 1 AEOD/DS P/ROAB 1 1 AEOD/DS P/TPAB 1 1 D AEOD/TTC 1 1 DEDRO 1 1 NRR/DORS/OEAB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRI L/RPEB 1 1 NRR/DRSS/PEPB 1 1 NRR/PMAS/ILPB1 1 1 NRR/PMAS/ILPB2 1 1 NUDOCS-ABSTRACT 1 1 OE DIR OGC/HDS2 RES/HFB 1

1 1

1 GN1

~LFILE 02 01 1

1 1

1 1

1 EXTERNAL EGGG/BRYCE i J H ~ ~ 1 1 NRC PDR 1 1 NSIC 1 1 D

D NOTE TO ALL "RIDS" RECIPIENTS:

D PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P l-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 24 ENCL 24

"=.'.. '.,: "- Pennsylvania Power & Light Company Two North Ninth Street ~Allentown, PA 18101-1179 ~ 215/774-5151 Robert G. Byram Senior Vice President-Nuclear 21 5/774-7502 Submiffed pursuant to 10CFR50.59 DEC 22 1993 Mr. Thomas T. Martin Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 SUSQUEHANNA STEAM ELECTRIC STATION 10CFR50.59

SUMMARY

REPORT - 1992 PLA<061 FILE R41-2A

Dear Mr. Martin:

Pursuant to 10CFR50.59(b), enclosed please find a summary report of the safety evaluations approved during the period from January 1, 1992 to December 31, 1992 for Susquehanna SES Units 1 and 2.

The report format is as follows:

SER No.- Unique number for each safety evaluation.

Cross Reference- Reference to the document for which the safety evaluation was prepared.

Description Change- A brief description of the change made to procedures, equipment or tests.

Summary- A summary of the three requirements for determining an unreviewed safety question as definedin 10CFR50.59(a) (2).

If you have any questions, please contact J. B. Wesner at (215) 774-7911.

Very truly yours, R.. B Enclosure

+ 70.g! t J 93123002722 93>>22 8 pDR ADOCW 05000 PDR R.

FILE R41-2A PLA4061 Mr. Thomas T. Martin cc: .~NRC Document Control Desk (original)

Mr. R. J. Clark, NRC Sr. Project Manager Mr. G. S. Barber, NRC Sr. Resident Inspector

SER NO.: 92-001 9312300272 CROSS

REFERENCE:

DCP 89-301B, Rev. 0, Unit ¹2 E

DESCRIPTION OF CHANGE:

Install a second isolation valve (double disc gate) upstream of the existing isolation globe valve in the feedwater heater tube side vent piping. Also, replace the existing isolation globe valves with new globe valves that are functionally equivalent to the existing globe valves.

SUMMARY

No. This modification does not jeopardize the function of any safety related equipment. The new valves interface with the feedwater system which is a non-Q system and not required for the safe shutdown of the plant. FSAR Section 10.4.7 and Chapter 15 were reviewed for this analysis.

No. This modification only interfaces with non-Q systems and does not create a possibility for an accident of a different type than previously evaluated in the FSAR. The installation and design will be in accordance with all original codes and standards.

No. This modification replaces valves in the feedwater system which is a non-safety related system and does not reduce the margin of safety as defined in the Basis of the Technical Specifications.

SER NO.: 92-002 CROSS

REFERENCE:

DCP ¹90-3108F, ILL, Unit ¹1 DESCRIPTION OF CHANCE:

This proposed activity will make the necessary modifications to existing pipe support/restraint configurations that are required in order to reduce the total number of mechanical snubbers located on selected SSES piping systems. The intended goal of this proposed action is to eliminate unnecessary snubbers which are currently installed on SSES piping.

SUMMARY

No. The systems included in this proposed modification were reanalyzed to justify the removal of unnecessary snubbers and their pipe support configurations are to be modified accordingly. The reanalysis is in accordance with the original design basis and will be performed in accordance with accepted industry codes. All aspects of the existing qualification calculations have been addressed, all original interface design parameters such as equipment allowables have been considered and all system design requirements as specified in the Design Specifications and FSAR have been addressed in the reanalysis effort. As in the existing piping analysis, the applicable code design limits have been met to ensure piping integrity and system function; FSAR Section 15.0.3.1, 15.0.3.5.

No. The accident events/causes in FSAR Chapter 15.0 that are applicable to this proposed action involve equipment malfunctions/failures and pipe breaks. The proposed action only reduces the number of seismic and hydrodynamic restraints (snubbers) on selected piping systems by reanalyzing each line using optimum restraint configurations.

No. A review of the following Technical Specification Sections for the proposed changes concluded that the safety functions of these systems would not be affected:

3/4.4.1, 3/4 4.9, 3/4.5.1, 3/4.5.2I 3/4.4.2, 3/4.4.7, 3/4.5.1, 3/4.7.3, 3/4.3.7, 3/4.4.8, 3/4 6 1 I 3/4 6 3 3/4 7 4

0 SER NO.: 92-003 CROSS

REFERENCE:

DCP ¹91-9062, Unit ¹1 DESCRIPTION OF CHANCE:

The purpose of this modification is to install new throttling valves which are designed for service within the operating range of the CRD system and are less susceptible to failure than the present valves.

SUMMARY

No. The failure and effects analysis for the control rod drive system as provided in FSAR Section 4.6 has been reviewed and determined to be unaffected.

The accident cases which apply to the control rod drive system are loss of coolant and radioactive release from subsystems and components. This modification is being performed between the control rod drive pump and the hydraulic control unit (HCU). As stated in FSAR Section 4.6.1.1.2.4.2, in order for a loss of coolant to occur due to a leak in this portion of the system, an additional failure of a hydraulic control unit check valve is required. Any loss of coolant due to a failure in piping being modified would be insignificant as compared to the volume evaluated in FSAR Sections 15.6.4, 16.6.5 and 1 5.6.6.

No. The control rod drive system is designed in accordance with G.E. Specification 22A7468 which requires that no single failure of a component, structure, or service shall prevent a reactor shutdown. This modification does not introduce any new failure modes. Although the ability to shutdown the plant without mechanical control rods is an off-design base scenario, FSAR Section 15A.6.6.3 discusses the plant's capability to do so through the manual actuation of the Standby Liquid Control System.

n,. No. Loss of the CRD pumps or a break in the supply piping can cause a loss of system flow/pressure which can result in the inability to move the control rods by normal means.

Extensive deterioration of accumulator charging line check valves could cause a sufficient number of accumulators to discharge and result in a loss of scram capability. Technical Specification Surveillance Requirements 4.1.3.5 verifies the operability of the accumulator check valves with the pumps shutdown at least every 18 months. This modification does not affect pump or system function or operation. The operability of the scram accumulator check valves is not diminished, hence, the margin of safety is not reduced.

SER NO.: 92-004 CROSS

REFERENCE:

SCP E91-2057 through 2060, Unit ¹1 DESCRIPTION OF CHANGE:

Change the fixed tap setting of existing Class 1E, 37.5 KVA, 480-208Y/120V Instrument AC Transformers from 468/120 to 456/1 20.

SUMMARY

No. The proposed instrument transformer tap change from 468/120 to 456/120, does not affect any of the postulated initiating events identified in Chapter 6, 8, 15 and NRC question 40.6 of the FSAR or NUREG-0776 because the change in tap does not create an unacceptable overvoltage condition and provides sufficient voltage to operate safety-related devices when required.

No. The proposed change ensures safety by raising the voltage levels at the Class 1E 120 VAC loads during all plant operating conditions. The proposed action provides sufficient voltage at the distribution panel and terminals of the affected circuits to allow the load devices to operate and meet their design requirements.

No. The proposed action maintains the margin of safety by allowing safety-related devices to operate properly even under degraded grid voltage conditions. The transformer tap change from 468/1 20 to 456/1 20 maintains the Class 1E, 120 VAC distribution panel bus voltage within the established design criteria stated. Technical Specification Sections 3.3.3, 3.3.7.5, 3.8.3.1, and 3.8.3.2 were reviewed to conclude that the margin of safety was not reduced.

SER NO.: 92-005 CROSS

REFERENCE:

SCP-]92-1001, Unit ¹ - Common DESCRIPTION OF CHANGE:

The purpose of the proposed setpoint change package is to change the trigger setpoints for the Seismic Monitoring System.

SUMMARY

No. The Seismic Monitoring System performs no automatic function, it provides indicating, recording and alarm functions only. This change meets the requirements of Regulatory Guide 1.12. The lower trigger setpoints ensure the initiation of arialysis at a sufficiently low ground acceleration level to encompass the anticipated design spectra for SSES below OBE and SSE levels. The proposed setpoint change does not affect the existing requirements.

No. The Seismic Monitoring System performs no automatic function, it provides indicating, recording and alarm functions only. The proposed setpoint change meets all the requirements specified by FSAR Section 3.7b.4.1, and 3.13, and Regulatory Guide 1.12.

No The new setpoints will be well within the range of the instrumentation. The ranges of the instrumentation are discussed in Table 3.3.7.2-1 of the Technical Specification.

The proposed change has no impact on acceptance limits, it meets the Regulatory Guide 1.12 requirements which are invoked by the Technical Specification 3.3.7.2 bases.

SER NO.: 92-006 CROSS

REFERENCE:

DCP 091-3012Z, Unit t2 DESCRIPTION OF CHANGE:

Replace the existing underrated fuses in Non-Class 1E battery chargers, motor control centers, uninterruptible power supplies, fuse box and load center within the Unit t2 Non-Class 1E 250 VDC system with fuses appropriately qualified for DC service.

SUMMARY

No. The function and operation of the connected loads and the 250 VDC subsystems supplying power to these loads as described in FSAR Section 8.3.2 are not modified by this action.

No. Installation of replacement fuses and fuse blocks does not alter the function or operation of the Non-Class 1E loads connected to the Class 1E buses. The modification does not affect the 250 VDC subsystems supplying power to the loads as described in FSAR Section 8.3.2.

No. The bases for Technical Specification 3/4.8.2 does not address margin of safety relative to protection of DC power sources. However, since the modification improves the operability of the power sources by correcting an existing deficiency of the power circuit fuses, the proposed action would not result in a reduction in the margin of safety of the DC power system.

SER NO.: 92-007 CROSS

REFERENCE:

DCP ff83-0151, Unit - Common DESCRIPTION OF CHANGE:

This modification enhances the operation of the Liquid Radwaste Filters (OF302 A & B), as described in FSAR Section 11.2.2.2. The modification replaces existing high and low level sensors with more reliable sensors, installs a redundant high level overfill control and installs a new analog level indication system which allows the operator to continuously monitor level during the fill process.

SUMMARY

No. The proposed modification does not impact any station design features that are used in radioactive release analysis for postulated radwaste system failures. All pressure boundary changes are designed and installed in accordance with the Group D (Augmented) quality assurance requirements. Therefore, no concerns are raised by this change over an increase in the probability of vessel or associated piping failure. Failure analysis for the proposed changes thus would be enveloped by the Accident Analysis, "Postulated Radioactive Releases due to Liquid Radwaste Tank Failure," (FSAR Section 15.7.3).

No. Worst case scenarios for any accident or malfunction of the radwaste system have been conservatively analyzed in the FSAR Section 15.7.3. The proposed changes do not involve a change in system operation or add a more severe type of failure mode which would have an effect on this evaluation.

No. Technical Specification 3.11.1.3 relates only to liquid waste system operability. This modification will provide an upgrading of the existing filter level instrumentation. These changes will not degrade current system operability. They are designed to enhance operability.

SER NO.: 92-008 CROSS

REFERENCE:

DCP ¹89-3006C, Unit ¹1 DESCRIPTION OF CHANCE:

Installation of a second isolation valve (double disc gate) upstream of existing vent or drain line isolation globe valve. The removal of the existing globe valve by cutting and capping the pipe.

SUMMARY

No. The new valves, located in the Turbine Building between elevations 656'nd 699',

interface with the feedwater system which is a non-Q system and not required for the safe shutdown of the plant. FSAR Section 10.4.7, 10.4.10, and Chapter 15 were reviewed for this analysis.

During operation, radioactive steam is present in the extraction steam piping and the feedwater heater shells. The valves being replaced are installed in piping which connects to these barriers. FSAR Section 15.6.4 address steam system piping breaks outside containment. The building fault event for breaks outside containment is the complete severance of one of the four main steam lines. The calculated exposure for this accident is illustrated in FSAR Table 15.6-9 and represents only a small fraction of the 10CFR 100 guidelines. This scenario envelops the same failure for the pipe being modified.

No. The replacement valves, and piping where applicable, will be welded into the existing system as are the original valves and piping in order to minimize the risk of leakage in accordance with FSAR Section 10.4.10.3. All components being replaced or added are of similar physical characteristics to the original components and, therefore, any type of postulated component failure will not change.

m. No. This modification is non-safety related and does not affect systems having Technical Specification requirements.

SER NO.: 92-009 CROSS

REFERENCE:

DCP ¹90-9032C, Unit ¹2 DESCRIPTION OF CHANCE:

At panel 2C007, cable NK2K0073E shall be made a spare cable. Rework internal wiring from terminal board AA to TRSH-C12-2R018.

At panel 2TC611-B, cable NK2K0073E shall be made a spare cable. Part of GE cable 8758/C12A-001 is made spare.

At panel 2C664, part of GE cable 8758/C12A-001 is made spare.

At panel 2C651, change annunciator window 05A, 5-8 to a blank window.

SUMMARY

No. FSAR Section 4.1.3 - Reactivity Control Systems, Section 4.6.1.1 - Control Rod Drive System Design, Section 4.6.1.1.2.6 - Instrumentation, FSAR Section 7.7.1.2 - Reactor Manual Control System - Instrumentation and Controls and FSAR Chapter 15.0- Accident Analysis were reviewed to determine the impact of this modification. The Chapter 15 analyses, Section 15.4- Reactivity and Power Distribution Anomalies discusses control rod related anomalies. FSAR Section 4.6.1.1.1.1.1 describes the control rod drive mechanical system safety design bases.

No. FSAR Section 4.1.3- Reactivity Control Systems, Section 4.6.11- Control Rod Drive System Design, Section 4.6.1.1.2.6- Instrumentation and FSAR Section 7.7.1.2 - Reactor Manual Control System (RMCS) - Instrumentation and Controls were reviewed for impact of the proposed change.

m. No. The Unit ¹2 Technical Specifications have been reviewed for impact of the proposed modification, specifically Sections 3.0 and 4.0- Limiting Conditions for Operation and Surveillance Requirements, Section 3/4.1 - Reactivity Control Systems, and Section B3/4-Bases for Sections 3.0 and 4.0.

SER NO.: 92-010 CROSS

REFERENCE:

DCP ¹91-3022, Unit ¹1 DESCRIPTION OF CHANCE:

Installation of a ring and beam assembly on Jet Pumps 1, 2, 10, 11, 12 and 20.

SUMMARY

No. This modification improves the securing of Jet Pump Sensing Line IJPSL) and reduces the probability of failure of the sensing line. Updated FSAR Sections 6.3, 15.6.2 and 15.6.3, NUREC-0776, Section 3.9.5 were reviewed.

No. This modification improves the securing of the JPSL, which helps to eliminate sensing line failure caused by high vibratory stresses due to resonant frequencies. The JPSL failure is already bounded by Instrument Line Break, UFSAR Section 15.6.2 and LOCA Analysis Section 6.3.

No. The Jet Pump Instrumentation System is required to determine the operability of the individual jet pumps as required by Technical Specification Section 3.4.1.2. Since this modification will not affect the operability of the jet pump, the margin of safety is not reduced. It may, however, require placing the unit in hot shutdown per Technical Specification 3.4.1.2, since jet pump operability cannot be verified.

SER NO.: 92-011 CROSS

REFERENCE:

SCP J-91-1040, Unit ¹2 DESCRIPTION OF CHANCE:

Reduction of the Unit ¹2 instrument air dryer settings to be identical to those of the Unit ¹1 pressure valves.

SUMMARY

No. The new setpoints would be below the 125 psig vessel design pressure identified in Table 9.3-1 of the FSAR. According to Table 3.2-1 of the FSAR, non-safety pressure vessels are protected in accordance with the requirements of the ASME Pressure Vessel Code, Section Vill. The reduced settings, therefore, would be within that allowed by the FSAR and the applicable portions of the referenced ASME Code. FSAR Section 9.3.1.1 states that Instrument Air is non-safety related, and that air operated components, essential for safe shutdown, are designed to fail safe.

No. FSAR Section 9.3.1.1 states that all safety-related components using instrument air are designed to fail safe upon a loss of air. NUREG 1275, Volume 2, Operating Experience Feedback Report - Air Systems Problems, documents cases where a partial loss of instrument air (header depressurization) has caused safety-related, air-supplied components to fail in the unsafe position.

No. The proposed action affects only the Instrument Air System and will, in no way, affect the ability of any safety-related component to perform its safety function.

SER NO.: 92-012 CROSS

REFERENCE:

TP-134-038, Unit 01 DESCRIPTION OF CHANGE:

This change allows tie-in of a temporary outage chiller to the Reactor Building Chilled Water System during a Unit 81 outage.

SUMMARY

No. Section 9.2.12.3, "Reactor Building Chilled Water" and 9.4.2, "Reactor Building Ventilation" were reviewed. Installation of this temporary system will not impact operation of Primary Containment Isolation, the only safety function of the RBCW system.

No. The installation of this temporary system does not adversely impact the operation or safety function of any existing installed equipment.

No. A review of the Technical Specification Sections 3.6.3, 3.6.1.7, 3.6.5.1 concluded that the margin of safety is not reduced.

SER NO.: 92-013 CROSS

REFERENCE:

DCP 090-3100, Unit (f2 DESCRIPTION OF CHANGE:

Removal of the valve body drain line (SP-DCA-250-1) and its associated isolation valves (243F0278, 243F0288) from reactor recirculation pump suction valve HV-243F0238.

SUMMARY

No. The applicable design basis accident in the SAR which was reviewed for potential impact by this change is described in FSAR Section 15.6, "Decrease in Reactor Coolant Inventory."

'he specific evaluation is described in FSAR Section 15.6.5, "Loss-of-Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary) - Inside Containment." The probability of failure of the drain piping has already been determined to be reduced and FSAR Section 3.6.2.1.4.6 specifically excludes high energy piping less than as equal to 1 inch NPS from postulated break requirements.

No. The change does not introduce any new modes of failure for the affected system, nor does it impact the system in such a manner that the probability of any type of accident would be increased.

No. This change does not affect the bases for the Technical Specification limits or programmatic requirements identified. The Technical Specification Sections 3/4.4.8 and 4.0.5, and FSAR Section 5.2.4 were reviewed.

SER NO.: 92-014 CROSS

REFERENCE:

SCP ¹E912053, Unit ¹1 DESCRIPTION OF CHANGE:

Changes the overload heater coils for motor-operated valves HV-11313 and HV-11314.

SUMMARY

No. This change improves the availability of the motor-operated valves. Since there is no change in the power circuit design function, the performance of its safety function as evaluated'in FSAR Sections 6.2.4 and 15.6.5 is not affected.

No. The proposed action increases the available motor terminal voltage during the worst case accident condition to provide sufficient motor starting torque to perform its safety function.

It also maintains motor protection during maintenance activities with the present overload heater sizing criteria. The FSAR Sections 6.2.4 and 15.6.5 were reviewed.

No. The containment isolation valves are covered by Technical Specification Section 3.6.3.

Table 3.6.3-1 details the maximum isolation time for the valve. No changes to the plant design are involved.

SER NO.: 92-015 CROSS

REFERENCE:

SCP ¹E912054, Unit ¹2 DESCRIPTION OF CHANGE:

Changes the overload heater coils for motor-operated valves HV-21313 and HV-21314.

SUMMARY

No. This change improves the availability of the motor operated valves. FSAR Sections 6.2.4 and 15.6.5 were referenced.

No. The proposed action increases the available motor terminal voltage during the worst case accident condition to provide sufficient motor starting torque to perform its safety function.

It also maintains motor protection during maintenance activities with the present overload heater sizing criteria. FSAR Sections 6.2.4 and 15.6.5 were reviewed.

No. The containment isolation valves are covered by Technical Specification Section 3.6.3.

Table 3.6.3-1 details the maximum isolation time for the valves. The proposed action does not involve any change to the plant design.

SER NO.: 92-016 C ROSS

REFERENCE:

OP-105-004, 005, Unit ¹1 DESCRIPTION OF CHANCE:

The proposed action is to maintain Reactor Building Zone I and III ventilation operable by temporarily powering selected safety-related Zone I/III ventilation damper schemes. These schemes would be normally disabled during the 4 KV Bus 1C (1A203) and Bus 1D (1A204) outages.

SUMMARY

No. This temporary power is provided from the same division, is reliable Class 1E power, and will be available during the BUS outages. There are no failure mode differences between the permanent and temporary power supplies. Based on the analysis, the temporary Class 1E 120 VAC supply is acceptable, since it does not degrade the Class 1E system below an acceptable level.

No. FSAR Section 7.1.2a.3 was reviewed with respect to connection of temporary power to Class 1E circuits. This temporary configuration is substituting the normal Class 1E power source with another source from the same division; therefore, no electrical separation violation exists. It has been evaluated and found to be an acceptable alternate power source.

No. The proposed installation of temporary Class 1E power supply does not prevent any design safety function of safety-related equipment. The analysis shows the utilization of temporary Class 1E power sources does not degrade the Zone I and III Reactor Building HVAC system below an acceptable level.

Technical Specification Section 3/4.6.5, "Secondary Containment" was reviewed; the proposed action does not reduce the margin of safety.

SER NO.: 92-017 C ROSS

REFERENCE:

QP-205-004, 005, Unit 42 DESCRIPTION OF CHANGE:

The proposed action is to maintain Reactor Building Zone II and III ventilation operable by temporarily powering selected safety-related Zone II/IIIventilation damper schemes. These schemes would be normally disabled during the 4 KV Bus 2C (2A203) and Bus 2D (2A204) outages.

SUMMARY

No. This temporary power is provided from the same division, is reliable Class 1E power, and will be available during the BUS outages. There are no failure mode differences between the permanent and temporary power sources. Based on the analysis, the temporary Class 1E 120 VAC power source is acceptable, since it does not degrade the Class 1E system below an acceptable level.

No. FSAR Section 7.1.2a.3 was reviewed with respect to connection of temporary power to Class 1E circuits. This temporary configuration is substituting the normal Class 1E power source with another source from the same division; therefore, no electrical separation violation exists.

No. The proposed installation of temporary Class 1E power supply does not prevent any design safety function of safety-related equipment. The analysis shows that utilization of temporary Class 1E power sources does not degrade the Zone II and III Reactor Building HVAC system below an acceptable level.

Technical Specification Section 3/4.6.5, "Secondary Containment" was reviewed; the proposed action does not reduce the margin of safety.

SER NO.: 92-018 C ROSS

REFERENCE:

DCP ¹91-9080Z, 91-9080A, 91-9080B, Unit ¹1 DESCRIPTION OF CHANGE:

This modification pertains to the Unit ¹1 LPCI outboard injection valves and will be issued as a segmented package. The changes will modify the A Loop Valve F015A, and the 8 Loop Valve F015B.

SUMMARY

No. This modification will be designed, constructed, and tested in accordance with the same Code and design basis as the existing plant configuration. The valves being modified are ASME Section III, Class 1 Valves and will be modified in accordance with a Code repair plan per Section XI of the ASME Code. The following FSAR Sections were reviewed:

5.2.5.3.2, 5.2.5.2, 3.11, 3.11.1A.1.

No. The valve is being modified in accordance with the original Code and design basis for the affected system and as such will not compromise its structural integrity. No new failure modes will be created. Furthermore, no new accidents or malfunctions have been created by this modification.

No. These valves have Technical Specification requirements which address the performance criteria for the valves in regard to their response to protect all three of the primary barriers.

They are Technical Specification requirement 4.5.1, Sections 3.4.3.2 and 3.6.12. After a review of these sections, it was concluded that the margin of safety would not be affected.

SER NO.: 92-019 CROSS

REFERENCE:

DCP ¹91-9077, Unit ¹1 DESCRIPTION OF CHANGE:

This modification will upgrade some of the pneumatic control components in the HVAC system to improve its reliability. It will reduce the alarm setpoint for high temperature in the MG Sets to provide a larger operator response time window. Finally, this modification will add an alternate damper positioning capability to permit the operator to place the system in a stable operating mode for the duration of the repairs.

SUMMARY

No. Section 15.3.1 of the FSAR evaluates the consequences of tripping both Reactor Recirculation Pumps due to a number of different causes, and classifies this as a moderate frequency event. This is the only accident related in any way to the functioning of the MG Set Ventilation System.

No. This modification does not alter the normal function of the ventilation system or the manner in which it provides cooling for the MG Sets.

The only component failure which was not previously considered is the new pressure regulator. The effects of its failure or malfunction, however, in its role as backup following the failure or malfunction of the temperature controller, is no different than the failure of the controller itself.

Operator error can result in mispositioning of the new block valves in the pneumatic signal lines to the positioners. The two normal alignments of these valves have one valve open and one closed. The misalignments would involve having both valves open or both closed.

m. No. There is no failure related to this modification that will affect the margin of safety of any component covered by the Technical Specifications.

SER NO.: 92-020 CROSS

REFERENCE:

DCP ¹90-3014, Rev. 1, Unit ¹1 DESCRIPTION OF CHANGE:

This modification installs a modified disc and wear strips in the Unit ¹1 outboard HPCI steam supply isolation valve.

SUMMARY

No. Single active component failures (SACF) such as failure of a valve to open or close are moderate frequency accidents as stated in FSAR Section 15A.3.3.2. Single failure of any single electrical device is similarly evaluated to have a moderate frequency. Failure of a valve to close whether it be from the wear strip binding the valve, a mechanical problem with the actuator, or an electrical supply failure to the operator, each have the same probability of occurrence as defined in FSAR Section 15.0.3 and result in the same consequence.

No. The failure of a valve to function, single active component failure, is evaluated for each of the accident scenarios as stated in FSAR Section 15.0.3.2.1. Single active component failures can render individual systems within the ECCS inoperable without affecting the reliability of the ECCS system as discussed in FSAR Section 6.3.2.5. Failure of individual ECCS systems and the alternate ECCS equipment available for safe shutdown is tabulated in FSAR Table 6.3-5. These measures will preclude a reactor cavity drain-down event and an accident of a different type will not be created.

No. The disc modification and addition of wear strips to valve HV155F003 does not affect the automatic isolation closure time of the valves as specified in Table 3.6.3-1 of the Technical Specifications.

SER NO.: 92-021 CROSS

REFERENCE:

DCP ¹90-3015, Unit ¹2 DESCRIPTION OF CHANGE:

This modification installs a modified disc and wear strips in the Unit ¹2 outboard HPCI steam supply isolation valve.

SUMMARY

No. Single active component failures (SACF) such as failure of a valve to open or close are moderate frequency accidents as stated in FSAR Section 15A.3.3.2. Single failure of any single electrical device is similarly evaluated to have a moderate frequency. Failure of the valve to close, whether it be from the wear strip binding the valve, a mechanical problem with the actuator, or an electrical supply failure to the operator, each have the same probability of occurrence as defined in FSAR Section 15.0.3 and result in the same consequence.

No. The failure of a valve to function, single active component failure, is evaluated for each of the accident scenarios as stated in FSAR Section 15.0.3.2.1. Single active component failures can render individual systems within the ECCS inoperable without affecting the reliability of the ECCS system as discussed in FSAR Section 6.3.2.5. Failure of individual ECCS systems and the alternate ECCS equipment available for safe shutdown is tabulated in FSAR Table 6.3-5.

These measures will preclude a reactor cavity drain-down event and an accident of a different type will not be created.

I I I. No. The disc modification and addition of wear strips to valve HV155F003 does not affect the automatic isolation closure time of the valves as specified in Table 3.6.3-1 of the Technical Specifications.

SER NO.: 92-022 CROSS

REFERENCE:

DCP ¹91-9082Z, 91-9082A, 91-9082B, Unit ¹1 DESCRIPTION OF CHANCE:

A hole will be drilled through the reactor side valve bridge or the reactor side valve disc to connect the bonnet cavity with the reactor side piping. This is being done to provide a pressure relief path between valve bonnet cavity and the reactor side system piping.

SUMMARY

No. This modification will be designed, constructed, and tested in accordance with the same Code and design basis as the existing plant configuration. The valves being modified are ASME Section III, Class 1 valves and will be modified in accordance with a Code repair plan per Section XI of the ASME Code. The following FSAR Sections were reviewed:

5.2.5.3.2, 5.2.5.2, 3.11, 3.11.1A.1.

No. The valve is being modified in accordance with the original Code and design bases for the affected system and as such will not compromise its structural integrity. No new failure modes will be created.

No. These valves have Technical Specification requirements which address the performance criteria for the valves in regard to their response to protect all three of the primary barriers.

They are Technical Specification requirement 4.5.1, Sections 3.4.3.2 and 3.6.12. After a review of these sections, it was concluded that the margin of safety would not be affected.

SER NO.: 92-023 CROSS

REFERENCE:

DCP ¹91-3025Z, 91-3026Z, 91-3027Z, 91-3028Z, Unit ¹1 DESCRIPTION OF CHANCE:

Rewire the Non-Class 1E computer inputs from the MOC switches to the upper auxiliary switch contacts.

Rewire the Class 1E MOC switch contact for the annunciator logic into the Class 1E control circuitry of the breaker. These are done to resolve all the issues associated with a welded closed MOC switch contact used for Non-Class 1E computer or annunciator inputs.

SUMMARY

No. The FSAR, including Chapters 6 and 15, and NUREG-0776 were reviewed. It was determined that the safety function of all the breakers is not changed by rewiring the computer inputs and the annunciator contact.

No. The FSAR, including Chapters 6 and 15, and Section 8.1.6.1.q were reviewed. The conclusion was that the rewiring, splicing and termination of field cables and internal wiring is in accordance with existing approved installation termination procedures.

No. The existing margin of safety for each of these systems is dependent on the operation of the buses and breakers. The proposed action associated with the breakers does not affect the operability requirements, surveillance requirements or any existing margin of safety in any Technical Specifications.

SER NO.: 92-024 CROSS

REFERENCE:

Procedure ¹TP-141-006, Unit ¹1 DESCRIPTION OF CHANGE:

Removal of calcium carbonate scale on Main Condenser.

SUMMARY

No. Review of the various design basis accidents identified in Section 15 of the FSAR reveal that none of these accidents are affected by this cleaning procedure since it does not impact the integrity, function, or performance of any plant safety-related equipment.

No. The Circulating Water System design basis is discussed in Chapter 10, Section 10.4.5 of the FSAR. The proposed tube cleaning process will not have any adverse effects on any materials in the system. The chemical product used for cleaning the condenser tubes will not adversely impact or compromise the life of any materials the product will contact.

There is no change to the design basis and there is no impact on the operation of any systems in the plant as a result of the cleaning operation.

No. The plant Technical Specific'ations have been reviewed. It was found that the Technical Specifications are not affected by this tube cleaning operation.

SER NO.: 92-025 CROSS

REFERENCE:

Procedure ¹TP-154-072, Unit ¹1 DESCRIPTION OF CHANGE:

This procedure will drain the A Loop ESW piping to support maintenance and/or modification work.

SUMMARY

No. FSAR Section 6.3.6 discusses NPSH margin for ECCS equipment in the event of a failure in an ECCS watertight pump room. With only one pump room flooded, adequate NPSH still exists for the remaining ECCS pumps. FSAR Section 9.2.5 discusses the capability of supplying the TBCCW and RBCCW heat exchangers with ESW. The TBCCW and RBCCW heat exchangers are non-essential heat exchangers and are not required to mitigate the consequences of an accident.

No. The TBCCW/RBCCW heat exchangers are non-essential heat exchangers and removing a backup cooling source will not affect safe shutdown of the plant.

No. Draining of Unit ¹1 ESW Loop A will be accomplished within Technical Specification LCO constraints. Operability of the A RHRSW system as required by Technical Specification 3.7.1.1 will not be affected by performance of this procedure.

SER NO.: 92-026 CROSS

REFERENCE:

Procedure ¹TP-154-073, Unit ¹1 DESCRIPTION OF CHANCE:

This procedure will drain the B Loop ESW piping to support maintenance and/or modification work.

SUMMARY

No. FSAR Section 9.2.5 discusses the capability of supplying the TBCCW and RBCCW heat exchangers with ESW. This procedure removes this capability for Unit ¹1 on the B Loop.

The TBCCW and RBCCW heat exchangers are non-essential heat exchangers and are not required to mitigate the consequences of an accident.

No. The TBCCW/RBCCW heat exchangers are nonessential heat exchangers and removing a backup cooling source will not affect safe shutdown of the plant.

No. Draining of Unit ¹1 ESW Loop B will be accomplished within Technical Specification LCO constraints. Operability of the B RHRSW system as required by Technical Specification 3.7.1.1 will not be affected by performance of this procedure.

SER NO.: 92-027 CROSS

REFERENCE:

DCP ¹91-3036, Unit ¹1 DESCRIPTION OF CHANCE:

Replacement of the yoke clamps with 3-way and 2-way block clamps on a portion of the final feedwater sample line in the steam tunnel.

SUMMARY

No. In accordance with FSAR Section 9.3.2.3, "The process sampling system is not required to function during an accident, nor is it required to prevent or mitigate the consequences of an accident."

This modification does not alter any of the sampling system design bases as discussed in FSAR Subsection 9.3.2.1. A break of this sample line is enveloped by the instrument line break as discussed in FSAR Section 15.6.2. However, failure of the sample line would require an unscheduled shutdown, potentially challenging safe shutdown systems.

Replacement of the yoke type tube clamps will reduce stresses in the sample line, thereby, reducing the potential for an unscheduled shutdown.

No. The process sampling system is described in FSAR Section 9.3.2 and is not required to function during an accident, nor is it required to prevent or mitigate the consequences of an accident. The main feedwater system is described in FSAR Section 10.7 and, likewise, it is not required to function during or after an accident.

Loss of sample line flow for any reason results in loss of chemistry monitoring until corrected but does not create any possibility for unevaluated accidents or malfunctions.

Likewise, pressure boundary failures of lines of this size have previously been evaluated and are enveloped by the instrument line break analysis as described in FSAR Section 1 5.6.2.

No. Loss of pressure boundary for the final feedwater sample line would result in the release of wet steam in the vicinity of the break. The inventory loss from the feedwater system from breakage of a 1/4" tube would automatically be made up by the feedwater control system. Because the break size of 0.15 inch diameter is significantly less than the 0.25 inch diameter orifice considered by the instrument line break analysis in FSAR Section 15.6.2, the established design margins are not reduced.

SER NO.: 92-028 CROSS

REFERENCE:

DCP ¹91-3037, Unit ¹2 DESCRIPTION OF CHANCE:

Replacement of the yoke clamps with 2-way block clamps.

SUMMARY

No. In accordance with FSAR Section 9.3.2.3, "The process sampling system is not required to function during an accident, nor is it required to prevent or mitigate the consequences of an accident."

This modification does not alter any of the sampling system design bases as discussed in FSAR Subsection 9.3.2.1. A break of this sample line is enveloped by the instrument line break as discussed in FSAR Section 15.6.2. However, failure of the sample line would require an unscheduled shutdown, potentially challenging safe shutdown systems.

Replacement of the yoke type tube clamps will reduce stresses in the sample line, thereby, reducing the potential for an unscheduled shutdown.

No. The process sampling system is described in FSAR Section 9.3.2 and is not required to function during an accident, nor is it required to prevent or mitigate the consequences of an accident. The main fe'edwater system is described in FSAR Section 10.7 and, likewise, it is not required to function during or after an accident.

Loss of sample line flow for any reason results in loss of chemistry monitoring until corrected but does not create any possibility for unevaluated accidents or malfunctions.

Likewise, pressure boundary failures of lines of this size have previously been evaluated and are enveloped by the instrument line break analysis as described in FSAR Section 1 5.6.2.

No. Loss of pressure boundary for the final feedwater sample line would result in the release of wet steam in the vicinity of the break. The inventory loss from the feedwater system from breakage of a 1/4" tube would automatically be made up by the feedwater control system. Because the break size of 0.15 inch diameter is significantly less than the 0.25 inch diameter orifice considered by the instrument line break analysis in FSAR Section 15.6.2, the established design margins are not reduced.

SER NO.: 92-029 CROSS

REFERENCE:

DCP 490-3101C, Unit C2 DESCRIPTION OF CHANCE:

Installation of a ball valve in the 3" common header between the 20" condensate effluent header and the first 2" seal water branch line (to RFP 2A).

SUMMARY

No. The FSAR has been reviewed; specifically, Chapter 15, "Accident Analysis" and Section 10.4.7, "Condensate and Feedwater."

FSAR Section 10.4.7.1 states the design basis of the condensate and feedwater systems is to return condensate from the condenser hotwell to the reactor at the required flows, pressure and temperature. Since this modification only adds a 3" ball valve in the seal water piping to the RFPs and will always be open when the system is in operation, no system function is altered and the design basis is not changed.

The consequences of this manual valve being inadvertently closed during operation is that the seal water low pressure alarms would sound in the control room, thus alerting the operator to take appropriate actions.

The Accident Analyses contained in Chapter 15 are unaffected by the proposed modification.

No. The addition of this modification will not change the normal operation of the condensate system or the feedwater system. It merely provides isolation capability between the condensate demineraiizer effluent and the seal water components to the reactor feed pumps (all 3 pumps).

No. This change adds a valve which is designed and built to the same standards, codes and design conditions as the condensate system. The new valve will not in any way degrade the operation of the condensate/feedwater system, or more specifically, the reactor feed pump seal water system.

SER NO.: 92-030 CROSS

REFERENCE:

DCP ¹92-9001, Unit ¹1 DESCRIPTION OF CHANGE:

Connection of spare conductors in parallel with conductors in existing cables to reduce voltage drop at 1C690A. Connection of spare conductors in parallel with conductors in existing cables and addition of a new cable to reduce voltage drop at 1C690B.

SUMMARY

No. The proposed connecting of spare conductors in parallel with active conductors and increasing the size of conductors does not affect any of the postulated initiating events identified in Chapter 6, 8 and 15 of the FSAR and NUREG-0776 because the changes provide sufficient voltage to operate the H,O, Analyzers when required.

No. The proposed action ensures safety by reducing the voltage drop to the H,O, Analyzers during all plant conditions. It also does not change the existing design of the H,O, Analyzers.

No. Since this change does not degrade operation of the H,O, Analyzers, even under degraded grid voltage conditions, it does not reduce the margin of safety of the systems as defined in the basis of Technical Specification Section: 3.3.7.5 Accident Monitoring Instrumentation 3.8.3.1 Power Distribution - Operating 3.8.3.2 Power Distribution - Shutdown SER NO.: 92-031 CROSS

REFERENCE:

DCP ¹91-3022, Unit ¹1 DESCRIPTION OF CHANGE:

Reducing the unsupported lengths of jet pump instrument sensing lines on six jet pumps (1, 2, 10, 11, 12 &

20).

SUMMARY

No. This modification improves the securing of the jet Pump Sensing Line gPSL) and reduces the probability of failure of the sensing line. Updated FSAR Sections 6.3, 15.6.2, and 15.6.3 were reviewed.

No. This modification improves the securing of the Jet Pump Sensing Line, which helps to eliminate sensing line failure caused by high vibratory stresses due to resonant frequencies.

The jPSL failure is already bounded by Instrument Line Break, FSAR Section 15.6.2.

No. The jet Pump Instrumentation System is required to determine the operability of the individual jet pumps as required by Technical Specification Section 3.4.1.2. The limiting condition for operation is that all jet pumps be operable. The bases for this section states that an inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does in the case of a design-basis-accident, increase the blowdown area and reduce the capability of ref looding the core. Thus, all jet pumps shall be operable at all times. Additionally, the failure of a sensing line, the likelihood of which is decreased by this modification, is not a safety problem since it would not effect jet pump operation. It may, however, require placing the unit in hot shutdown per Technical" Specification 3.4.1.2 since jet pump operability cannot be verified.

SER NO.: 92-032 CROSS

REFERENCE:

DCP ¹91-9004Z, 91-9004A through 91-9004M, Unit ¹1 DESCRIPTION OF CHANCE:

Replacing 12 of the 24 breakers in Unit ¹1, 125 VDC, Class 1E distribution panels 1D614, 1D624, 1D634, 1D644 because they are oversized. They are replaced with breakers having trip ratings/characteristics which are compatible with cable derated ampacities, greater than connected load currents and coordinated with downstream subfuses.

SUMMARY

No. The replacement breakers are dedicated replacements for original equipment with trip ratings which match the derated ampacities of the connected cables in order to minimize to the extent practicable the effects of a sustained limited magnitude fault. The failure mode and effects analysis for the replacement breakers is bounded by existing analysis provided in FSAR Table 8.3-21.

No. Installation of properly sized and dedication tested replacement breakers and properly sized and qualified fuses does not modify the functions of the system loads. The proposed modification does not introduce any failure modes different than already analyzed in FSAR Table 8.3-21.

No. This change limits, to the extent practical, damage to adjacent cables of other systems in the same raceway.

SER NO.: 92-033 CROSS

REFERENCE:

DCP ¹91-9068A through 91-9068X, Unit ¹1 DESCRIPTION OF CHANCE:

There are 37 of 72 breakers that will be replaced in Unit ¹1, 125 VDC, Non-1E distribution panels 1D615, 1D625, 1D635, and 1D645 which are oversized. They will be replaced with breakers having trip ratings/characteristics which are compatible with cable ampacities, greater than connected load currents and coordinated downstream subfuses.

SUMMARY

No. The replacement breakers are identical replacements for original equipment except that the trip ratings match the derated ampacities of the connected cables in order to minimize to the extent practicable the effects of a sustained limited magnitude fault.

No. Installation of properly sized replacement breakers and properly sized fuses does not modify the non-safety related functions of the system loads. The proposed modification does not introduce any failure modes different than already analyzed in FSAR Table 8.3-21.

No. This change limits, to extent practical, damage to adjacent cables of other non-safety systems in the same raceway.

SER NO.: 92-034 CROSS

REFERENCE:

DCP ¹91-9012F through 91-9012W, Unit ¹2 DESCRIPTION OF CHANCE:

There are 22 of 25 breakers in Unit ¹2, 125 VDC, Non-1E distribution panels 2D615, 2D625, 2D635 and 2D645 that will be replaced because they are oversized. They will be replaced with breakers having trip ratings/ characteristics which are compatible with cable derated ampacities, greater than connected load currents and coordinated with downstream subfuses.

SUMMARY

No. The replacement breakers are identical replacements for original equipment except that the trip ratings match the derated ampacities of the connected cables in order to minimize to the extent practicable the effects of a sustained limited magnitude fault.

No. Installation of properly sized replacement breakers and properly sized fuses does not modify the non-safety related functions of the system loads. The proposed modification does not introduce any failure modes different than already analyzed in FSAR Table 8.3-21.

No. This change limits, to extent practical, damage to adjacent cables of other non-safety systems in the same raceway.

SER NO.: 92-035 CROSS

REFERENCE:

Unit ¹1: OP-139-001, PCAF 1-91-1194 Unit ¹2: OP-239-001, PCAF 2-92-0047 DESCRIPTION OF CHANCE:

A procedure change to OP-1(2)39-001 to allow flow through Unit ¹1 condensate demineralizers in excess of 5760 gpm.

SUMMARY

No. Section 15.6.6 of the FSAR describes a feedwater line break outside containment. Failure of a condensate demineralizer pressure boundary would produce an accident which is similar to this analyzed event. However, the increased flow does not increase the probability of failure of the pressure boundary. All components which experience increased flow can accommodate the increased flow without exceeding their design limits for structural integrity.

No. Chapter 15 of the FSAR was reviewed. No accident will be created which is not already addressed. No failures are expected which could cause an accident.

No. The Basis Section of the Technical Specification was reviewed. The margin of safety is not reduced as defined in the basis. Condensate demineralizer flows are not mentioned in the basis.

SER NO.: 92-036 CROSS

REFERENCE:

OP-105-004, 005; Unit 01 DESCRIPTION OF CHANGE:

The change is to maintain Reactor Building Zone I & II ventilation OPERABLE by temporarily powering selected safety related Zone IfZone III ventilation damper schemes. These schemes are normally disabled.

SUMMARY

No. This temporary power is provided from the same division, is reliable Class 1E power, and will be available during the BUS outages. There are no failure mode differences between the permanent and temporary power supplies.

No. FSAR Section 7.1.2a.3 was reviewed with respect to connection of temporary power to Class 1E circuits. This temporary configuration is substituting the normal Class 1E power source with another source from the same division.

No. The applicable Technical Specification Section is 3/4.6.5, "Secondary Containment." The basis of the Technical Specification is to minimize any ground level releases of radioactive material which may result from an accident.

The proposed installation of temporary Class 1E power supply does not prevent any design safety function of safety-related equipment. The power supply has been analyzed and shows the utilization of temporary Class 1E power sources does not degrade the Zone I

& III Reactor Building HVAC system below an acceptable level.

SER NO.: 92-037 CROSS

REFERENCE:

DCP ¹91-9073, Unit ¹1 DESCRIPTION OF CHANCE:

This modification relocates "A" and "C" switches from the stanchions to the adjacent missile shield wall.

Mounting the "A" and "C" switches to the wall will reduce slightly the level of vibration because the wall is more rigid than the stanchion.

SUMMARY

No. The design basis accident to be reviewed for potential impact by the modification is the increase in reactor pressure caused by loss of condenser vacuum, FSAR Section 15.2.5.

This event is categorized as an incident of moderate frequency.

Radiological consequences of this event as stated in FSAR Section 15.2.5.5 does not result in fuel failure; however, it does result in the discharge of normal coolant activity into the suppression pool via SRV operation. This event does not result in an uncontrolled release to the environment or exposure to plant personnel.

The existing installation for Unit ¹1 consists of four MCLVPS's being the same manufacturer and model (Barksdale). This modification replaces the "A" and "C" switches with a different manufacturer (SOR, Inc.). To introduce diversity, one switch from each division ("A" or "C", Division 1, and "B" or "D" from Division 2) would have to be different. Therefore, diversity is not introduced and per FSAR Section 7.3.1.1a.2 4.1.13.5, diversity of the MCLVPS's is not required.

No. Relocation and replacement of the "A" and "C" switches with the Nuclear Qualified SOR, Inc. vacuum switches does not alter the function or operation of the switches. The modification will demonstrate by qualification and calculation to operate without contact chatter as exhibited in the previous installation.

No. Technical Specification Section 3/4.3.2, "Isolation, Actuation, Instrumentation", ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of reactor systems.

Except for the MSIVs, the safety analysis does not address individual sensor response times.

Per Technical Specification Table 3.3.2-3.3.e, the response time for Condenser Vacuum Low is not applicable.

SER NO.: 92-038 CROSS

REFERENCE:

DCP ¹91-9084, Unit ¹1 DESCRIPTION OF CHANCE:

Installation of a backup scheme to prevent the potential transformer (PT) control fuse failure from causing an unnecessary line loss and a resultant Unit ¹1 reactor SCRAM.

SUMMARY

No. The proposed action affects only the carrier relaying scheme and the backup phase distance relaying scheme which provide transmission line protection in the event of an electrical fault on the power system. Failure of this relaying scheme to operate will not prevent the plant electrical system from performing its design safety functions. FSAR Chapters 6 and 15 were reviewed.

No. The proposed relay installation will be analyzed in order to satisfy II over I safety impact concerns. All required changes will be done solely in panel OC658 in the Main Control Room. This modification will be engineered according to existing standards and practices.

Electrical separation will be maintained according to existing requirements. No safety-related circuits will be affected. The proposed action increases the reliability of the carrier relaying scheme and the backup phase distance relaying scheme by providing fault protection consistent with their intended design and function. Redundancy of the existing relaying schemes will be maintained in that only one of the three protective relaying schemes is required to operate and clear a fault on the 230 KV line between the Unit ¹1 Synchronizing Breaker and the 230 KV switchyard. The proposed changes do not involve changes in system operation or add a more severe type of failure mode.

No. The proposed modification does not interact with required sources of offsite power nor with any Class 1E power supplies or equipment needed for safe shutdown or control of accident conditions.

Technical Specification Basis 3/4.8.1, 3/4.8.2 and 3/4.8.3 were reviewed.

SER NO.: 92-039 CROSS

REFERENCE:

SCP ¹E912053, Unit ¹1 DESCRIPTION OF CHANCE:

Replacement of the overload heater coils for motor-operated valves HV-11313 and MV-11314.

SUMMARY

No. The availability of the motor-operated valves is increased by ensuring that the available motor terminal voltage is sufficient for the valve to perform its safety function. This was previously evaluated in the FSAR Sections 6.2.4 and 15.6.5.

No. The available motor terminal voltage is increased during the worst case accident condition to provide sufficient motor starting torque to perform its safety function. It also maintains motor protection during maintenance activities with the present overload heater sizing criteria. FSAR Sections 6.2A and 15.6.5 were reviewed.

No. The containment isolation valves are covered by Technical Specification Section 3.6.3.

Table 3.6.3-1 details the maximum isolation time for the valves. No plant design change is involved.

SER NO.: 92-040 CROSS

REFERENCE:

DCP ff90-9104, Unit ff1 DESCRIPTION OF CHANCE:

Splitting the different functions of Scheme 2C1026 by separately fusing the circulating water pump switchgear breaker auxiliary relays supplied from power supply 2D625-06. This allows control power to now be available to start or stop any circulating water pump in the normal mode on any failure of the switchgear breaker auxiliary relays.

SUMMARY

No. The function of the Circulating Water System control power circuit is not changed or affected by this modification, which proposes the installation of an additional relay, indicating lights and fuses.

Chapters 6, 8, 10, 15 and NRC Questions of the FSAR and NUREG-0776 were reviewed.

No. The design basis and function of the DC Power System as analyzed in FSAR Section 8.3.2 and NUREG-0776 Section 8.3.2 is not altered by this modification.

The design basis and function of the Circulating Water System or its annunciation system as analyzed in FSAR Section 10 4.5 is not altered by this modification.

No. The new relay and three indicating lights require a DC power supply from the 125 VDC subsystems whose operability is governed by Technical Specifications Section 3/4.8.2. The addition of a DC load does not degrade the designed performance of the 125 VDC subsystems and does not reduce the margin of safety as defined in the basis of Technical Specifications Section 3/4.8.2.

The Circulating Water System has no safety-related functions as defined by FSAR Section 10.4.5.

SER NO.: 92-041 CROSS

REFERENCE:

DCP 90-3101A, Unit ff2 DESCRIPTION OF CHANCE:

Installation of a second isolation valve on elevation 656'f the Turbine Building between the recycle isolation valves HV-21615A through G, and the sluice/recycle/drain header of the condensate demineralizer system.

SUMMARY

No. The accident analysis contained in Chapter 15 are unaffected by the proposed modification.

Section 9.2.10 does not specifically discuss the Condensate Demineralizer System and, therefore, does not need to be addressed in this section.

Section 10.4.6.1 states that the design basis of the Condensate Cleanup System is to maintain condensate purity by removing various contaminants. This section also states the not-to-exceed effluent quality for the demineralizers. The design basis for the system is not affected by the proposed action.

No. In this modification, there is no equipment important to safety and no interface with safety equipment or systems.

No. This modification is non-safety related and does not affect systems having Technical Specification requirements. This change adds valves which are designed and build to the same standards and design conditions as the condensate demineralizer system. The new valves will not in any way degrade the operation of the condensate demineralizer system.

SER NO.: 92-042 CROSS

REFERENCE:

90-3101B, Unit ¹2 DESCRIPTION OF CHANGE:

Installation of a 3" ball valve upstream of the condensate demineralizer drain valve HV-21618 to provide isolation of the butterfly valve when maintenance is required.

SUMMARY

No. Since this modification only adds a 3" ball valve to a drain line, and does not alter the system function as described in FSAR 10.4.6.1, the design basis for the system is not affected by the proposed action.

FSAR Section 10.4.6.2.4 "Waste Systems" deals specifically with the regeneration waste discharges, one of which is the low conductivity, low solids wastewater to the Turbine Building sump where our new valve is located. Adding a manual isolation valve upstream of the existing air operated valve creates no adverse change to system operation or function.

The possibility of a leak caused by the addition of this new valve has been evaluated against the Accident Analysis in Chapter 15 of the FSAR, specifically Section 15.7.2.1-Miscellaneous Small Releases Outside Containment. Since the ball valve used for this application will comply with the original ANSI piping code, there will be no decrease of the design integrity of the system.

No. The addition of this modification will not change the normal operation of the condensate demineralizer system. It merely supplies redundant isolation capability to the systems

,n No.

drain line to LRW.

Technical Specification 3/4.11.1 sets limits to unrestricted areas for liquid radwaste treatment system operability in minimizing radioactive material in liquid waste prior to discharge. The additional isolation valve in the condensate demineralizer drain does not increase or add to the radioactive material limits, however, it may reduce the volume of liquid radwaste that must be processed, thus enhance system performance and reliability.

SER NO.: 92-043 CROSS

REFERENCE:

OP-068-122, OP-069-012, OP-069-050, Unit - Common DESCRIPTION OF CHANGE:

Bypass for processing chemical radwaste to the river after appropriate treatment through the atmospheric demineralizer.

SUMMARY

No. The consequences of an accident as previously evaluated in FSAR 15.7.3 or 2.4.13.3, are not increased. Temporary hosing, fittings, and valves meet design requirements and the new processing mode does not conflict with system line design pressures or temperatures.

No. The only failure that could be postulated as a result of the bypass and new processing mode would be an integrity failure resulting in draining the affected components, Distillate Sample Tank OT321 or Evaporator Concentrate Storage Tank OT322 and associated pumps, piping, and valving. This failure has been previously evaluated in the FSAR. FSAR Section 15.7.3, Postulated Radioactive Releases due to Liquid Radwaste Tank Failure, analyzes the complete release of activity from the concentrate's waste tank and bounds all liquid radwaste failures.

No. The following Technical Specifications are applicable to release of liquid effluents and remain unaffected: 3/4.3.7.10, 3/4.11.1.1, 3/4.11.1.2, 3/4.11.1.3, and 3/4.11.4.

SER NO.: 92-044 CROSS

REFERENCE:

TP-159-013, Unit ff1 DESCRIPTION OF CHANCE:

This test will perform a 45 psig LLRT of the MSIVs in the accident direction.

SUMMARY

No. FSAR Sections 5.4.5.4; Table 3.2.1; 9.1.4.2.5.2; Table 6.2.12; Table 6.2.22; 6.2.6.3 have been reviewed.

Revision 42 of the FSAR incorporated all requirements of 45 psig MSIV testing. All components within the vacuum (-30" H20 maximum) test volume have been evaluated for structural integrity. The safety function of the MSIVs and the Leakage Control System (LCS), to isolate with their appropriate containment isolation signal, will not be degraded with the relatively small vacuum being applied inside the test volume, and is not required to be operable during this test.

No. The MSL plugs will be pressurized below the evaluated pressure. All requirements for use of the MSL plugs are incorporated into the prerequisites and flowpath of TP-159-013. All other aspects of the testing are as described in the FSAR.

No. The basis section for Technical Specification 3/4.6.1.2 describes MSIV testing as an exemption to 10CFR50 Appendix] and that the "special requirements" are based on experienced degradation of the leak tightness of the valves. The "special requirements" mean the increased test frequency that is assigned to the valves, lower test pressure, and separate leakage criteria. TP-159-013 will be performed at the same frequency (18 months) and acceptance criteria (46 scfh, 22 slm). TP-159-013 will increase the test pressure to 45 psig but this change will highlight actual valve seating flaws and decrease maintenance on minimal defects, which can introduce valve problems.

SER NO.: 92-045 CROSS

REFERENCE:

Bypass 1-92-006, Unit ¹1 DESCRIPTION OF CHANCE:

Use of a station bypass to allow the main turbine bypass valves to open in a different sequential manner than stated in FSAR 10 4.4.2.

SUMMARY

No. The proposed action does not affect the ability of the Main Turbine Bypass Valve System to pass the required 25% steam flow as stated in FSAR Section 10.4.4.2. The proposed'ction only changes the opening sequence of the valves. The proposed action also does not affect the response opening time nor the ability to fast open on load reject. The Condenser is unaffected because, even though the sequence of opening is changed, the required steam flow is not. The Main Turbine Bypass Valve system will perform as required by the FSAR.

No. The original intent of the design of the Bypass Valve System to bypass 25% of the steam flow is still being met with the exception of the opening sequence. The intent of this evaluation is to provide a one-time operation in this manner to ensure that a BPV with questionable reliability is not knowingly used to bring the reactor to cold shutdown. The inherent problem with the bypass valve will be corrected during the Unit ¹1 6RFIO. On the next restart, the system will be returned to its original operating mode.

No. The Technical Specification Section 3/4.3.8 and Bases 3/4.7.8 which require valve cycling and a specific response time are not affected. The MCPR penalty is not required as a result of this bypass.

SER NO.: 92-046 CROSS

REFERENCE:

RE-081-030, Units ¹1 & ¹2 DESCRIPTION OF CHANCE:

Analysis of addition of Nuclear Energy Services (NES) LPRM Bender to Rev. 9 of RE-081-030.

SUMMARY

No. A Fuel Handling Accident as evaluated in FSAR 15.7.4 and 15A (Event 41), is the only postulated accident which is applicable to this procedural change. The radiological consequences of dropping a fuel assembly are well within the guidelines of 10CFR100.

The effects of a dropped LPRM Bender/LPRM are within the scope of a dropped bundle.

No. The LPRM is clamped into the LPRM Bender utilizing an air/

hydraulic power unit. Failure of the hydraulic cylinder will not drop the LPRM.

Should the LPRM break during the bending process, half of the LPRM would fall into the cavity/RPV area (the other half would remain in the LPRM benders clamp). An examination of the break area would then be performed, and retrieval of the fallen piece(s) would commence.

The use of the LPRM Bender will not violate PP&L's 5-foot shielding requirement for LPRM "hot ends".

No. From Technical Specification 3/4.9.7, Crane Travel -Spent Fuel Pool, LCO, loads in excess of 1,100 pounds shall be prohibited from travel over fuel assemblies in the spent fuel storage racks. Since the combined weight of an LPRM Bender and LPRM is only 550 pounds, this procedural addition does not impact the acceptance limit.

SER NO.: 92-047 CROSS

REFERENCE:

DCP 90-3035C, Unit ¹1 DESCRIPTION OF CHANCE:

Replacement of existing RWCU Pump 1P221A with a new sealless pump.

SUMMARY

No. FSAR Sections 5.4.8.1, 5.4.8.2 and 5.4.8.3 provide the criteria, system description and system evaluation for the Reactor Water Cleanup System. FSAR Tables 3.2-1 and 5.4-2 provide Reactor Water Cleanup System Equipment Design Basis and Data. FSAR Section 7.3.1.1A.2 4.1.9 addresses high flow isolation instrumentation.

These sections and tables and the design basis were reviewed for applicability. The proposed modification to replace the existing pumps does not alter the function of the Reactor Water Cleanup System.

No. The performance, function, classification and the flow capacity of the replacement pumps are the same as or better than that of the original pumps. These modifications do not introduce any new failure modes, nor do they introduce new single failures.

No. Technical Specification Sections 3.3.2, 4.3.2.1, 4.3.2.2 and 4.3.2.3 are applicable to Reactor Water Cleanup System. Table 3.3.2-2 lists the isolation actuation, instrumentation trip setpoints for the system. These sections and their implied basis were examined and it was determined that the margin of safety as defined in this basis is not affected by this modification.

SER NO.: 92-048 CROSS

REFERENCE:

DCP 91-3022, Unit ¹1 DESCRIPTION OF CHANCE:

Installation of six (6) clamps on jet Pump Sensing Line to prevent the instrument sensing lines from failing.

SUMMARY

No. FSAR Section 5.4.1 4 specifies that the jet Pump Assembly provides a floodable volume of 2/3 core height; however, this work is being performed with the core offloaded and therefore the core height is zero.

The FSAR Chapter 6 and 15 events, evaluate situations when fuel is in the reactor. Since no fuel will be in the vessel when this activity will be performed, there are no negative effects on safety.

The FSAR Section 7.7.1 discussion of refueling interlocks, especially the rod block generated when the monorail hoist is loaded and over the vessel, describes their function as preventing inadvertent criticalities. Since the core will be offloaded during this activity and no fuel or control blades will be handled, this rod block is not required. Also, the 1,000 pound overload cutout will be retained.

No. The core will be defueled during the activity, no new failure modes are created, and the plant will be in an improved condition after the modification is installed.

No. The Jet Pump Operability Technical Specification 3/4.4.1.2 is only required during Conditions 1 and 2. The Technical Specification 3/4.9.6 rod block is only required when handling fuel or control blades, which will not be done during this activity, and the 1,000 pound overload cutout will be retained. Since all this work is being completed with the core defueled, no decrease in any margin of safety will occur.

SER NO.: 92-049 CROSS

REFERENCE:

DCP 92-9021, Unit ¹1 DESCRIPTION OF CHANCE:

Addition of bolted splice plates to 4" vertical pipe member to provide access to Valve HV-151F047A during overhauls.

SUMMARY

No. The modification will be designed and constructed to the same criteria applicable to the original member.

No. The modification, including the added splice plates, is designed to accommodate the maximum calculated loading, and is designed in accordance with the same design criteria that is applicable to the original configuration.

No. There are no Technical Specifications applicable to the Reactor Building structural steel, of which this member is an integral part; and this modification does not affect the margin of safety of any other licensing basis documents.

SER NO.: 92-050 CROSS

REFERENCE:

SCP J92-2002, Unit ¹1 DESCRIPTION OF CHANCE:

Lower SRM downscale setpoint from 5.4 cps to 23.40 cps to facilitate outage functional testing.

SUMMARY

No. FSAR Section 15.4.1, 15.4.9, and Appendix A, Events 16 and 40, cover rod withdrawal errors and rod drop accidents. These accident scenarios are analyzed with an initial power of 0% rated. Therefore, the results of the FSAR transient analyses will still be valid with the lower SRM downscale setpoint.

No. Lowering the downscale setpoint does not introduce the possibility of an unanalyzed operating condition or accident. The safety analyses of FSAR Section 4.3 concerning criticality during refuel operations are not affected.

No. Unit ¹1 Technical Specification Table 3.3.6-2, Entry 3.d, states that the minimum downscale setpoint for the SRMs will be R0.7 cps if the signal to noise ratio (signal to noise ratio) is >2, or M3 cps if the SNR (2. An SNR 22 has not been confirmed at SSES, so the limit of 3 cps is used. The proposed downscale setpoint is more conservative than that in Table 3.3.6-2, and has taken into consideration actual instrument drift.

The lower downscale setpoint satisfies the criteria established in the Technical Specification. Section B 3/4.3.7.6 of the Technical Specification state that the SRM will provide the operators with flux information at very low power levels during start-up and shutdown. The proposed setpoint satisfies this criteria.

SER NO.: 92-051 CROSS

REFERENCE:

SCP f92-2003, Unit 02 DESCRIPTION OF CHANCE:

Lower SRM downscale setpoint from 4.1 cps to 23.40 cps to facilitate outage functional testing.

SUMMARY

No. FSAR Sections 15.4.1, 15.4.9, and Appendix A, Events 16 and 40, cover rod withdrawal errors and rod drop accidents. These accident scenarios are analyzed with an initial power of 0% of rated. Therefore, the results of the FSAR transient analyses will still be valid with the lower SRM downscale setpoint.

No. Lowering the downscale setpoint does not introduce the possibility of an unanalyzed operating condition or accident. The safety analyses of FSAR Section 4.3 concerning criticality during refuel operations are not affected.

No. Unit 42 Technical Specification Table 3.3.6-2, Entry 3.d, states that the minimum downscale setpoint for the SRMs will be h0.7 cps if the signal to noise ratio (SNR) is h2, or c3 cps if the SNR (2. An SNR R2 has not been confirmed at SSES, so the limit of 3 cps is used. The proposed downscale setpoint is more conservative than that in Table 3.3.6-2, and has taken into consideration actual instrument drift.

The lower downscale setpoint satisfies the criteria established in the Technical Specification. Section B 3/4.3.7.6 of the Technical Specification state that the SRM will provide the operators with flux information at very low power levels during start-up and shutdown. The proposed setpoint satisfies this criteria.

SER NO.: 92-052 CROSS

REFERENCE:

DCP 92-9012, Unit tl DESCRIPTION OF CHANGE:

Add upper and lower louvers on one (1) or both sides of all eight (8) Reactor Protection System (RPS) EPA enclosures to promote cooling and ventilation.

SUMMARY

No. The proposed action increases the ventilation to the EPAs in accordance with GE Design Specification 22A5941 which already indicates that cooling the EPAs by means of enclosure louvers is acceptable.

Failure of the louvers to perform properly (i.e., louver blockage), would not prevent the RPS equipment from performing its design safety functions. FSAR Chapters 6 and 15 were reviewed in support of this change.

No. This modification will be engineered in accordance with existing standards and procedures. No safety-related circuits will be affected. The proposed action increases the reliability of the RPS system by providing EPA ventilation consistent with their intended design and function. The proposed modification does not involve changes in system operation nor add a more severe or different type of failure mode. Loss of both RPS power supplies results in a reactor SCRAM, which ensures a safe condition for the plant, and initiates the NSSSS isolations to ensure containment integrity.

No. Technical Specification Basis 3/4.8.1, 3/4.8.2 and 3/4.8.3, "A.C. Sources, D.C.

Sources and Onsite Power Distribution" states, "the operability of the A.C. and D.C.

power sources and associated distribution systems during operation ensures that sufficient power will be available for one (1) safe shutdown of the facility, and (2) mitigation and control of accident conditions within the facility." (The RPS power supply is not a required power source per Technical Specifications 3/4.8.1, 8.2, 8.3).

The proposed modification will not impair the operation of any equipment or power supplies needed for safe shutdown or control of accident conditions. Technical Specification 3/4.8.4.3 specifies the operability and surveillance requirements of the EPAs. The proposed modification does not adversely impact EPA operability or surveillance criteria.

SER NO.: 92-053 CROSS

REFERENCE:

DCP 92-9022, Unit ¹1 DESCRIPTION OF CHANCE:

Install reworked disc arm in Valves HV-151F050A8 8.

SUMMARY

No. FSAR Chapters 15 (Accident Analysis), 6 and 3 have been reviewed for possible consequences concerning this modification and none were found. Both the new style LH disc arm replacement in Valve ¹HV-151F050A and the RH disc arm (converted LH disc arm) replacement in Valve ¹HV-151F0508 are designed and constructed to meet all manufacturer's specifications as an acceptable replacement disc arm.

No. This modification will replace the disc arms in both valves with disc arms of equivalent quality that are certified by the manufacturer as acceptable replacement items. The new disc arms will provide an improved seating of these check valves, thus allowing them to better perform their functions, and the modified LH disc arm will provide closure capability equivalent to the original design.

No. The Technical Specification applicable to these check valves is Section 3/4.6.3, "Primary Containment Isolation Valves" and Section 3/4.6.1, "Containment Systems" in particular, Subsection 3.6.1.2, "Primary Containment Leakage Rates."

The modification, including the machining of a LH disc arm for use as a RH disc arm, will provide Valves ¹HV-151F050A&B with a manufacturer certified and approved disc arm design that surpasses the original design. They will provide equivalent closure and improved seating capabilities, thus allowing the valves to accomplish their functions better.

SER NO.: 92-054 CROSS

REFERENCE:

TP-037-005, Units C1 & 02 DESCRIPTION OF CHANGE:

Cross-tie demineralizer water to condensate transfer to supply ECCS discharge piping keepfill.

SUMMARY

No. FSAR Section 6.3.2.2.5, Discharge Line Fill System states that: "The discharge line fillsystem consists of fill lines which provide a continuous supply of condensate from the condensate transfer system to the high points of the ECCS discharge piping.

Following initial venting and system fill, a pressure above atmospheric pressure is maintained at the system's high points to prevent air accumulation."

The Demineralizer Water System via the cross-tie on elevation 818'ill provide adequate quality water above atmospheric pressure for ECCS Keepfill. Should keepfill pressure not remain above atmospheric in each ECCS discharge piping high point, Technical Specification 3/4.5 ECCS Action Statements must be implemented which places each unit in a safe shutdown condition.

~ I No. The proposed evolution (TP-037-005) involves the possibility of losing ECCS Keepfill Pressure. The criteria for determining adequate ECCS Keepfill is discussed in Section 6.3.2.2.5 of the FSAR and Technical Specifications Section 3/4.5 details appropriate actions to take if adequate keepfill pressure is not maintained.

No. The basis for Technical Specification 3/4.5.1 states that for each loop of ECCS, "The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment." The proposed action will maintain a positive pressure and therefore will maintain the ECCS pump discharge piping full of water.

If the discharge piping cannot be maintained full of water, Technical Specification 3/4.5, Emergency Core Cooling Systems details the actions required after declaring HPCI, RHR, and Core Spray inoperable. The time allowed for reaching cold shutdown depends on the number of safety systems that are inoperable.

Since condensate transfer and demineralizer water systems are both non-Q and non-1E, there is no greater risk from this standpoint and therefore the margin of safety is not decreased.

SER NO.: 92-055 CROSS

REFERENCE:

DCP 92-3011, Unit 41 DESCRIPTION OF CHANGE:

Replacement of the worm, worm gear, motor pinion, and worm gear shaft on HV-155F002 to increase motor pullout torque capability allowing maximum torque switch setting to be increased.

SUMMARY

No. The modification will not impact the pressure retaining capability of the valve and therefore will not increase the probability of occurrence or the consequences of any decrease in reactor coolant accident previously analyzed in Section 15.6 of the FSAR.

The modification has no impact on the dynamic qualification of the MOV as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations.

The proposed modification has no impact on the ability of HPCI to perform its design intended function as described in FSAR Section 6.3.2.2.1 when called upon to mitigate the consequences of an accident analyzed in Chapter 15 of the FSAR.

No. The proposed change does not adversely impact the ability of the MOV to isolate against conditions associated with a HELB, does not increase the stroke time beyond that previously contained in the Design Basis for SSES (FSAR Table 6.2-1 2), does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis, and finally, does not affect any other equipment.

m. No. The bases for Technical Specification 3/4.3.2 and 3/4.3.3 discuss reactor system isolation actuation instrumentation and ECCS system actuation instrumentation. Since the modification affects no equipment other than HV-155F002, these bases are unaffected. The modification has no effect on HPCI operation and therefore does not necessitate a change to the basis for Technical Specification 3/4.5 "Emergency Core Cooling Systems." With regard to the containment isolation function of HV-155F002, the integrity of the valve will not be adversely affected and the isolation time remains within the accident analysis bounds of 50 seconds; therefore, the bases for Technical Specification 3/4.6.1 "Primary Containment" and Technical Specification 3/4.6.3 "Primary Containment Isolation Valves" are unaffected.

SER NO.: 92-056 CROSS

REFERENCE:

EDR G20018, Unit 41 DESCRIPTION OF CHANCE:

Demonstration that continuing with the existing configuration for operation on 3rd stage extraction steam piping tee connections until U1-7RIO does not represent an unreviewed safety question.

SUMMARY

No. The extraction steam system piping is covered in FSAR Section 10.4.10.1. As stated in Section il, the existing configuration is unchanged from what was installed in the plant originally. No appreciable erosion has been found in the areas of concern.

FSAR Section 15.1.1 and 15A.6.3.3 analyze the effect of loss of feedwater heating on the safe operation of the plant for decreases in heating capacity up to 100'F. Failure of the tee connection would result in the partial or full loss of extraction steam heating to the feedwater heater. This loss of a feedwater heater is evaluated in FSAR Section 15.1.1 to be bounded by an event which would incur a loss of up to 100'F in feedwater heating capacity.

Failure of the piping could result in a loss of condenser vacuum. Loss of condenser vacuum is evaluated in FSAR Section 15A.6.3 as a moderate frequency incident.

Alarms annunciate in the control room to indicate a condenser low vacuum condition and automatic main turbine trip and main steam line isolation circuitry is provided to protect the turbine and condenser from damage.

No. The extraction steam piping in the non-engineered configuration represents a reliability issue over the long term. In the short term, it functions as intended and does not create the possibility of different type of accident than previously evaluated.

No. The extraction steam system is not addressed in the Technical Specifications.

SER NO.: 92-057 CROSS

REFERENCE:

DCP 92-9007, Unit 01 DESCRIPTION OF CHANCE:

Addition of time delay into the high system flow portion of the RWCU leakage detection actuation logic to permit isolation valves to ride through the short duration high system flow transients.

SUMMARY

No. The addition of the time delay does not affect any of the postulated initiating events identified in FSAR Chapters 6 and 15.

No. Chapter 6, 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREC-0776 were reviewed to determine if the proposed action had the potential of creating a postulated initiating event which was not within the spectrum of events which transient or anticipated operational occurrences and accident conditions were analyzed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type.

The effects of the 60-second time delay in the high system flow portion of the RWCU leakage detection actuation logic were evaluated and did not identify any new malfunctions.

No. The proposed action does not affect the operability requirements of Section 3/4.6.3 or the maximum isolation times of Table 3.6.3-1.

The proposed action does not affect the Level 2 response time for the RWCU System Isolation as governed by Section 3/4.3.2, Table 3.3.2-3.

SER NO.: 92-058 CROSS

REFERENCE:

DCP 92-9024, Unit ¹1 DESCRIPTION OF CHANGE:

Weld overlay of DLA-102-1 due to erosion/corrosion.

SUMMARY

No. The applicable design basis accidents in the SAR which were reviewed for potential impact by this DCP are described in FSAR Sections 15.6 and 15A.6.3.3.

This change will not increase the probability of occurrence of a "Loss-of-Coolant" or "Loss of AII Feedwater Flow" accident since all standards of design, material and construction will be maintained for the modified piping.

No. The change does not introduce any new modes of failure for the affected system or components. The overlay is provided to add strength to the piping segment to reduce the probability of failure. Use of the qualified welding procedure precludes the weld process from having significant detrimental effects upon the material properties of the existing pipe or the weld overlay.

No. Technical Specification limits which are applicable are Section 3/4.4.8 as being the ASME Section XI ISI Program.

The surveillance requirement for verifying the structural adequacy of the reactor coolant system as described in FSAR Section 5.2.4 and Technical Specification 4.0.5 is the periodic inspection and pressure testing of the pressure retaining components of the reactor coolant pressure boundary, in accordance with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code and 10CFR50.

This change does not affect the bases for the Technical Specification limits or programmatic requirements identified. There are no changes in the inspection and pressure test boundaries due to this change since the weld overlay overlaps an existing pipe weld. Baseline NDE is to be performed in accordance with ASME Section XI requirements.

SER NO.: 92-059 CROSS

REFERENCE:

SCP j-91-1011 and SCP J-91-1012, Units ¹1 5 ¹2 DESCRIPTION OF CHANGE:

Decrease the setpoints of the recirculation pump motor bearing temperature high alarm to provide earlier indication of bearing high temperature.

SUMMARY

No. FSAR Sections 5.4.1 and 7.6.2a.8 describe the Recirculation Water System safety functions and system requirements. FSAR Sections 15.3.1, 15.3.2, 15.3.3, 15.4.4 and 15.4.5 describe Design Basis Accidents directly associated with the Recirculation Pumps. Review of these sections did not reveal any design features or safety functions impacted by this action.

No. This action conservatively reduces the high temperature alarm setpoint for the Recirculation Pump Motor Bearings. The action does not change the system design or function. The temperature switches have no safety function and provide alarm signals only. They do not provide trip or actuation signals to other equipment.

Reducing the setpoint will improve motor bearing protection by providing earlier warning of potentially damaging high bearing temperature.

No. Technical Specifications Sections 3/4.3.4, 3/4.4.1 and 3/4.10.4 were reviewed for applicability. None of these sections are affected by the action.

SER NO.: 92-060 CROSS

REFERENCE:

NL-91-030, Units ¹1 8 ¹2 DESCRIPTION OF CHANGE:

FSAR Change Notice 1610 to upgrade the MSIV closure capability description in the FSAR to be consistent with the valve's actual capabilities.

SUMMARY

No. FSAR Chapters 6.2, 6.3 and 15.7 were reviewed. The proposed action does not increase the consequences of an accident or malfunction evaluated in these sections.

In addition, the changes do not increase the probability of malfunction (valve configuration and capability has not changed) and do not affect the valve's ability to meet it's originally defined design safety basis as defined in the FSAR Section 5.4.5.1.

No. The proposed action does not create a possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR because no hardware changes are proposed and the operational/maintenance requirements are being made more conservative.

No. The bases for Section 3/4.4.7 Main Steam Isolation Valves, and 3/4.6.3 Primary Containment Isolation Valves, were reviewed. The specified maximum stroke times are to limit release of radioactive materials from the containment in an accident, and'or 3/4.4.7 to limit loss-ofwore coolant inventory. The consequences of an increased isolation time have been analyzed and are found to have no adverse impact on containment pressure, ECCS performance, or off-site dose predictions.

The MSIVs also provide an RPS input. For events which affect the MSIV stroke timing (LOCA), there is no affect on RPS actuation timing because the LOCA signaI trips RPS directly. For non-LOCA events, the MSIV timing is unaffected.

SER NO.: 92-061 CROSS

REFERENCE:

DCP 92-9013, Unit 82 DESCRIPTION OF CHANCE:

Add upper and lower louvers on one or both sides of all eight EPA enclosures to promote cooling and ventilation.

SUMMARY

No. The proposed action increases the ventilation to the EPAs in accordance with GE Design Specification 22A5941 which already indicates that cooling the EPAs by means of enclosure louvers is acceptable.

Failure of the louvers to perform properly (e.g., louver blockage) would not prevent the RPS equipment from performing its design safety functions. FSAR Chapters 6 and 15 were reviewed in support of this change.

No. This modification will be engineered in accordance with existing standards and procedures. No safety related circuits will be affected. The proposed action increases the reliability of the RPS system by providing EPA ventilation consistent with their intended design and function. The proposed modification does not involve changes in system operation nor add a more severe type of failure mode.

Loss of both RPS power supplies results in a reactor SCRAM, which ensures a safe condition for the plant, and initiates the NSSSS isolations to ensure containment integrity.

No. Technical Specification Basis 3/4.8.1, 3/4.8.2 and 3/4.8.3, "A.C. Sources, D.C.

Sources and Onsite Power Distribution" states, "the operability of the A.C. and D.C.

power sources and associated distribution systems during operation ensures that sufficient power will be available for (1) safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility." (The RPS power supply is not a required power source per Technical Specifications 3/4.8.1, 8.2, 8.3).

The proposed modification will not impair the operation of any equipment or power supplies needed for safe shutdown or control of accident conditions. Technical Specification 3/4.8.4.3 specifies the operability and surveillance requirements of the EPAs. The proposed modification does not adversely impact EPA operability or surveillance criteria.

SER NO.: 92-062 CROSS

REFERENCE:

DCP 90-3097, Unit P2 DESCRIPTION OF CHANCE:

Replacement of existing wedge gate valves in extraction steam start-up drains, FW heater start-up vents and drain cooler start-up vents with parallel slide gate valves to improve operation.

SUMMARY

No. The start-up vent valves and the extraction steam bypass drain valves being replaced by this modification are designed to fail open, as are the existing valves. The effect on plant operation of such a failure is unchanged by this modification.

The replacement valves are being purchased, installed, and tested in accordance with the same design codes as the original valves. The pipe hangers have been evaluated for suitability and/or adjustment in accordance with the original design criteria and the requirements of ANSI B31.1. The valves and associated piping, where applicable, are being welded into the existing system using approved procedures in order to minimize the chance of leakage in accordance with FSAR Section 10.4.10.3.

No. The replacement valves will perform the same function as the existing valves. The operators are sized to cycle the valves at full system pressure. The spring loaded parallel slide disc design effectively relies on the low pressure side seating surface for leak-tightness. The valves have been leak tested by the vendor to assure minimal seat bypass leakage. The in-service seat leakage for the new valves is expected to be lower than for the existing valves.

No. This modification does not affect systems having Technical Specification requirements. The new valves and piping do not change the function or operation of the system in which they are to be installed.

SER NO.: 92-063 CROSS

REFERENCE:

DCP 90-3108F, I&L, Unit 01 DESCRIPTION OF CHANCE:

Modifications to existing pipe support/restraint configurations to reduce the total number of mechanical snubbers.

SUMMARY

No. The applicable code design limits have been met to ensure piping integrity and system function as in FSAR Sections 15.0.3.1 and 15.0.3.5.

No. The proposed action only reduces the number of seismic and hydrodynamic restraints on selected piping systems by reanalyzing each line using optimum restraint configurations. The original allowable equipment interface loads are still being met and the pipe break criteria remains unchanged.

No. The margins of safety, as defined by the following Technical Specifications, will not be reduced.

3/4.4.1 Recirculation System, 3/4.4.9 RMR, 3/4.5.1 and 3/4.5.2 ECCS, 3/4.4.2 Safety/Relief Valves, 3/4.4.7 MSIVs, 3/4.7.3 RCIC, 3/4.3.7 Monitoring Instrumentation, 3/4.4.8 Structural Integrity, 3/4.6.1 Primary Containment, 3/4.7.4 Snubbers.

SER NO.: 92-064 CROSS

REFERENCE:

ON-037-001, Common DESCRIPTION OF CHANGE:

Establishment of a procedure to determine the cause of condensate transfer system loss, and actions to mitigate the impact to plant operations.

SUMMARY

No. FSAR Section 5.2.3 was reviewed. FSAR Sections 15.7.2.1, 15.6.4 and 15.6.6 were also reviewed. The proposed action does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.

No. FSAR Sections 15.7.2.1 and 6.3.2.2.5 bound the proposed action.

No. The following Technical Specification bases were reviewed:

3/4.4.9 Residual Heat Removal 3/4.5.1 and 3/4.5.2 ECCS - Operating and Shutdown 3/4.9.11 Refueling Operations - Residual Heat Removal and Coolant Circulation These bases document reasons for operability and requirements of ECCS during power, shutdown, and refueling operations.

Bases for 3/4.5.1 and 3/4.5.2 state pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

Actions prescribed by this procedure mitigate loss of ECCS/RCIC keep-fill by providing alternate means to maintain or restore it. Therefore, actions prescribed by this procedure actually provide additional margin to safety by minimizing the time ECCS/RCIC would be without keep-fill.

SER NO.: 92-065 CROSS

REFERENCE:

DCP 90-3108E, j5 K, Unit ¹1 DESCRIPTION OF CHANGE:

Modifications to existing pipe support/restraint configurations to reduce the total number of mechanical snubbers.

SUMMARY

No. The applicable code design limits have been met to ensure piping integrity and system function as in FSAR Sections 15.0.3.1 and 15.0.3.5.

No. The proposed action only reduces the number of seismic and hydrodynamic restraints on selected piping systems by reanalyzing each line using optimum restraint configurations. The original allowable equipment interface loads are still being met and the pipe break criteria remains unchanged.

No. The margins of safety, as defined by the following Technical Specifications, will not be reduced.

3/4.8.1 A.C. Sources, 3/4.3.1 RPS Instrumentation, 3/4.3.2 Isolation Actuation Instrumentation, 3/4.3.3 ECCS Actuation Instrumentation, 3/4.3.5 RCIC Actuation Instrumentation, 3/4.5.1 ECCS, 3/4.4.8 Structural Integrity, 3/4.6.1 Primary Containment, 3/4.7.4 Snubbers.

SER NO.: 92-066 CROSS

REFERENCE:

DCP 90-3058F, Unit ¹1 DESCRIPTION OF CHANCE:

Replace existing RBCW chiller condenser temperature control valve with new control valves to improve chiller condenser recirculation loop flow control for Chiller 1K206A.

SUMMARY

No. As discussed in FSAR Section 9.2.12.3, the only safety-related functions of the RBCW system are performed by the primary containment piping penetrations and the containment isolation valves.

The proposed modification adds no new function to the RBCW system as described in FSAR Section 9.2.12.3.

No. The proposed modification will install new temperature control valves and their temperature controller which do not alter the intended function of the SW system and the RBCW system as described in FSAR Section 9.2.1.1 and 9.2.12.3. The existing method of the service water temperature control by recirculation of the Service Water as described in FSAR Section 9.2.1.2 will be unchanged.

No. The RBCW system and the SW system operability does not contribute to any margin of safety as defined in the basis of any Technical Specification.. SSES Technical Specifications (Unit ¹1) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2, and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

SER NO.: 92-067 CROSS

REFERENCE:

DCP 90-3064F, Unit ¹1 DESCRIPTION OF CHANCE:

Replace existing RBCW chiller condenser temperature control valve with new control valves to improve chiller condenser recirculation loop flow control for Chiller 1K206B.,

SUMMARY

No. As discussed in FSAR Section 9.2.12.3, the only safety-related functions of the RBCW system are performed by the primary containment piping penetrations and the containment isolation valves.

The proposed modification adds no new function to the RBCW system as described in FSAR Section 9.2.12.3.

No. The proposed modification will install new temperature control valves and their temperature controller which do not alter the intended function of the SW system and the RBCW system as described in FSAR Section 9.2.1.1 and 9.2.12.3. The existing method of the service water temperature control by recirculation of the service water as described in FSAR Section 9.2.1.2 will be unchanged.

No. The RBCW system and the SW system operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit ¹1) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2, and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

SER NO.: 92-068 CROSS

REFERENCE:

OP-172(272)-001, Units ¹1 & ¹2 DESCRIPTION OF CHANGE:

Establishing condenser vacuum with the secondary steam jet only.

SUMMARY

No. FSAR Secti'on 11.3.2.3.4 states that (the Offgas) water removal skid and guard bed may be operated in parallel during start-up to accommodate the increased offgas flow of up to 300 SCFM (consisting mostly of air).

The proposed procedural steps are similar to a start-up in that a high flow will be seen. The duration of the high flow will be somewhat longer than that expected for a normal start-up. As previously noted, the ability for the various condensers to drain will be the limiting factor until the main condenser vacuum is adequate to provide the driving force for the condensate drains. Most of the water that would be carried from the Motive Steam Jet Condenser and/or Recombiner condenser would be collected in the offgas delay piping drains which eventually route this condensate to LRW.

No. The proposed action, evacuation of the main condenser using the secondary steam jet air ejector via the recombiner and charcoal absorber systems, will in no way alter or affect the assumptions made in the FSAR analysis of Section 15.7.1 and 15.7.1.11.1. Each of the above potential failure modes, is still the most limiting and is unaffected by the proposed action. This action will in all cases stay within the bounds of the FSAR analysis.

No. Technical Specifications 3.11.2.4 and 3.11.2.6 address the requirements for the Gaseous Radwaste Treatment System. Technical Specification 3.11.2.4 dictates the circumstances under which the offgas system must be operable while Technical Specification 3.11.2.6 specifies Hydrogen and Oxygen limits for operation. The Offgas System and Secondary Air Ejector System (SJAE) shall be in service using auxiliary steam as the prime mover, which is in compliance with Technical Specification Section 3.11.2.4. Due to greater than normal flows expected in this evolution, the H2 analyzers will be jumpered out (per the procedure) and requires chemistry sample taken every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per Table 3.3.7.11.-1 of the Technical Specification.

SER NO.: 92-069 CROSS

REFERENCE:

DCP 91-8029Z, 91-3030Z, 91-3031Z and 91-3032Z, Unit ¹2 DESCRIPTION OF CHANGE:

Modifications to all Unit ¹2 4.16 KV breakers to prevent welded closed computer or annunciator MOC switch contacts as a result of a failure in the non-Class 1E system.

SUMMARY

No. The FSAR, including Chapters 6 and 15, and NUREG-0776 were reviewed to determine if the proposed action has an effect on the spectrum of postulated initiating events for which transients or anticipated operational occurrences and accident conditions were analyzed. The rewiring of the computer inputs from the MOC switches to the auxiliary switch and the rewiring of the annunciator contact into the Class 1E portion of the breaker control circuit does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR or NUREG-0776.

No. As a result of review of Chapters 6 and 15 and FSAR Section 8.1.6.1.q, the rewiring of the computer inputs from the MOC switches to the auxiliary switch and the rewiring of the annunciator contact into the Class 1E portion of the breaker control circuit does not create a possibility of an accident or malfunction of a different type.

No. The operability of every system governed by Technical Specifications requires the operability of the 4.16 KV safety-related buses 2A201, 2A202, 2A203 and 2A204 and their breakers. The existing margin of safety for each of these systems is dependent on the operation of the buses and breakers. The proposed action associated with the breakers does not affect the operability requirements, surveillance requirements or any existing margin of safety in any Technical Specification. In fact, the proposed action ensures breaker and system operability for failures in the non-Class 1E computer or annunciator systems.

SER NO.: 92-070 CROSS

REFERENCE:

DCP 91-3033A&B, Unit ¹1 and Common DESCRIPTION OF CHANGE:

Addition of corrosion monitoring probes in ESW system.

SUMMARY

No. The installation of the corrosion probes do not affect the overall performance and operation of the ESW system because the probe assembly is a dead-leg length of piping with the probe tip minimally protruding into the flow path. The existing system piping has been reanalyzed to confirm that the added weight of the probe assembly will not over stress the piping during a seismic event. FSAR Sections reviewed are 3.6, 6.2, 6.3, 9.2.5, 1.2.2.8.3, 3.1.2.1.5 and Chapter 15.

No. Based on the description and safety evaluation of the ESW system provided in FSAR Section 9.2.5, the installation of the corrosion probes does not create a different type of accident from those described in FSAR Sections 3.6, 6.2, 6.3 and Chapter 15.

A new failure mode for the ESW system has not been created.

No. The bases of the Technical Specifications have been reviewed for the Emergency Service Water System (3/4.7.1.2). Also reviewed was Regulatory Guide 1.29 (Seismic Design) and General Design Criteria 5 (Sharing of Safety-Related Systems) and 45 (Inspection of Cooling Water Systems). The corrosion probes, which chiefly alter the pressure boundary, do not affect either system cooling capacity, system redundancy or system ability to withstand a single failure.

SER NO.: 92-071 CROSS

REFERENCE:

DCP 91-3034A&B, Unit ¹2 and Common DESCRIPTION OF CHANGE:

Addition of corrosion monitoring probes in RHR system.

SUMMARY

No. The installation of the corrosion probes do not affect the overall performance and operation of the RHRSW system because the probe assembly is a dead-leg length of piping with the probe tip minimally protruding into the flow path. The existing system piping has been evaluated to confirm that the added weight of the probe assembly will not over stress the main process piping during a seismic event. FSAR Sections reviewed are 3.6, 6.2, 6.3, 9.2.6, 1.2.2.8.2, 3.1.2.1.5 and Chapter 15.

No. Based on the description and safety evaluation of the RHRSW system provided in FSAR Section 9.2.6, the installation of the corrosion probes does not create a different type of accident from those described in FSAR Sections 3.6, 6.2, 6.3 and Chapter 15.

A new failure mode for the RHRSW system has not been created.

No. The bases of the Technical Specifications have been reviewed for the RHR Service Water System (3/4.7.1.1). Also reviewed was Regulatory Guide 1.29 (Seismic Design) and General Design Criteria 5 (Sharing of Safety-Related Systems) and 45 (Inspection of Cooling Water Systems). The corrosion probes, which chiefly alter the pressure boundary, do not affect either system cooling capacity, system redundancy or system ability to withstand a single failure.

SER NO.: 92-072 CROSS

REFERENCE:

SCP J-921020, Unit ¹1 DESCRIPTION OF CHANCE:

Decrease the RBCCW heat exchanger discharge header low pressure alarm setpoint from 78 psig to 71 psig to alleviate a nuisance alarm in the main control room.

SUMMARY

No. The RBCCW system is discussed in FSAR Section 9.2.2, Reactor Building Closed Cooling Water System. Per Section 9.2.2.3, the RBCCW system has no safety-related function and it's failure will not compromise any safety-related system or component or prevent a safe shutdown of the plant. The affected pressure switch does not cause any automatic actions.

No. The device affected by the proposed setpoint change cannot affect the input conditions for accidents previously evaluated in the FSAR.

No. The RBCCW system has no associated Technical Specifications. Changing the setpoint of PSL-11308 will not affect the safety limits or limiting safety settings of equipment required for safety as described in the Technical Specifications.

SER NO.: 92-073 CROSS

REFERENCE:

DCP 91-3020, Unit ¹2 DESCRIPTION OF CHANGE:

Modification of RCIC steam admission valve to improve speed control of the RCIC turbine in order to improve the overall reliability of the RCIC system.

SUMMARY

No. A total failure of the steam admission valve or its control circuit would not result in an accident as previously evaluated in the SAR.

The equipment being added to the RCIC system is highly reliable and the low probability of occurrence of a malfunction of this equipment is more than offset by the increased system reliability that will be gained by having a smooth start-up of the RCIC turbine and not challenging the overspeed trip.

The FSAR Chapter 15 accident analyses were reviewed for impact but none was found. RCIC may be used to help maintain reactor water level in case of a LOCA, however, as stated earlier per FSAR Section 7.4.2.1.2.1.8, RCIC meets the single-failure criterion on a network basis with HPCI. It is not necessary for RCIC alone to meet the single-failure criterion, in itself, since its function is duplicated or backed up by other systems. A malfunction of the modified steam admission valve or its modified control circuit'would have the same consequences as a malfunction in the present design.

No. The new electrical components (auxiliary relays and time delay relay) being installed by this modification are highly reliable. However, should one of these components fail, the only operating scenario that might occur which is not similar to what is seen today is that the steam admission valve might remain stuck at about 10% open leaving the turbine turning at idle speed. This low probability event is enveloped by failure of the RCIC system as discussed in FSAR Section 5.4.6.2.5.4.

No. RCIC system operability is addressed in the Technical Specifications in the Plant Systems Section 3/4.7.3 and Bases Section 3/4.7.3. This modification has no impact on RCIC operability as is described in these sections.

In addition, RCIC system actuation instrumentation is addressed in the Technical Specifications in Instrumentation Section 3/4.3.5 and Bases Section 3/4.3.5. This modification does not involve or impact, in any way, this instrumentation.

SER NO.: 92-074 CROSS

REFERENCE:

SCP E-921019 DESCRIPTION OF CHANGE:

Increase the CO, discharge time settings for control structure north, center, and south cable chases to 218, 171, and 230 seconds respectively to ensure a 50% concentration of CO, is maintained in the chases for a minimum of 15 minutes.

SUMMARY

No. The low pressure carbon dioxide system is discussed in FSAR Section 9.5.1.2.6 and in Revision 3 of the Fire Protection Review Report.

The proposed increase in discharge amounts for the cable chase systems will not make them the largest CO, systems, therefore, the CO, storage tank design basis is unaffected. The CO, systems are not inputs to the accidents discussed in Chapter 15 of the FSAR.

No. The CO, systems do not affect the initial conditions for accidents previously evaluated in Chapter 15 of the FSAR. The increased quantity of CO, to be discharged will not significantly affect the components in the cable chases.

No. The low pressure carbon dioxide fire protection systems are discussed in Unit ¹1 and Unit ¹2 Technical Specification Section 3.7.6.3. The basis for the CO, storage tank design is unaffected, hence, the minimum tank level and pressure is unaffected. The CO, system does not affect the limiting safety settings of equipment required for safety.

SER NO.: 92-075 CROSS

REFERENCE:

DCP 91-3021, Unit ¹2 DESCRIPTION OF CHANGE:

Reduce the unsupported lengths of jet pump instrument sensing lines on six jet pumps to prevent the recirculation pump vane passing frequency from reaching resonant frequency.

SUMMARY

No. This modification improves the securing of the jPSL and reduces the probability of failure of the sensing line. Therefore, the modification does not increase the probability of occurrence of an accident as previously evaluated in the SAR.

(Updated FSAR Sections 6.3, 15.6.2 and 15.6.3, NUREG-0776, Section 3.9.5).

The modification does not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the SAR, because the failure of the sensing line has been decreased by this modification.

No. The HPSL failure is already bounded by Instrument Line Break, UFSAR Section 15.6.2 and LOCA Analysis Section 6.3.

No. The Jet Pump Instrumentation System is required to determine the operability of the individual jet pumps as required by Technical Specification Section 3.4.1.2. The limiting condition for operation is that all jet pumps are operable. The bases for this section states that an inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does in the case of a design-basis-accident, increase the blowdown area and reduce the capability of ref looding the core. Thus, all jet pumps shall be operable at all times.

Since this modification will not effect the operability of the jet pump for the reason stated previously in this SER, the margin of safety is not reduced.

SER NO.: 92-076 CROSS

REFERENCE:

SCP E91-2063 and E91-2064, Unit ¹2 DESCRIPTION OF CHANGE:

Change voltage tap setting from 468/120 to 456/120 for the Class 1E, 37.5 KVA, 408-208Y/120V, Instrument AC Transformers 2X216 and 2X226 to improve terminal voltage at 120 VAC safety-related loads with the degraded grid voltage of 0.912 pu at the Class 1E 4160 VAC buses.

SUMMARY

No. This change does not impact the analysis in FSAR Chapters 6, 8, 15, NRC Question 40.6 of the FSAR or NUREG-0776.

No. The transformer tap change will not result in any misoperation in the Class 1E 120 VAC system, under normal, DBE or plant shutdown conditions.

No features of changing the taps have been identified that would indicate the existence of any mechanism for creating a malfunction of a different type than any previously analyzed.

No. The proposed action maintains the margin of safety by allowing safety-related devices to operate properly even under degraded grid voltage conditions. The transformers (2X216 and 2X226) tap change from 468/1 20 to 456/1 20 maintains the Class 1E, 120 VAC distribution panels (2Y216 and 2Y226) bus voltages within the established design criteria.

The Class 1E 120 VAC load devices rated less than 120 VAC nominal have been analyzed to show that these load devices operate properly with 120 VAC distribution panel voltages developed with the instrument transformer tap set at 456/120.

The proposed action does not reduce the margin of safety of the systems as defined in the bases of Technical Specification Sections 3.3.3, 3.3.7.5, 3.8.3.1 and 3.8.3.2.

SER NO.: 92-077 CROSS

REFERENCE:

TP-164-019, Unit ¹1 DESCRIPTION OF CHANGE:

Cleaning of M-G Set Hydraulic Fluid Coolers

SUMMARY

No. Review of the various design basis accidents identified in Section 15 of the FSAR reveal that none of these accidents are affected by this cleaning procedure since it does not impact the integrity, function or performance of any plant safety-related equipment.

This procedure does not affect any of the design parameters of any safety-related components or the function and performance of any safety-related systems.

No. The Service Water System design basis is discussed in Chapter 9, Section 9.2.1 of the FSAR. The proposed cleaning process will not have any adverse effects on any materials in the system.

There is no change to the design basis and there is no impact on the operation of any systems in the plant as a result of the cleaning operation.

The Motor-Generator Sets are discussed in Chapter 5, Section 5.4.1, Reactor Recirculation Pum s. The proposed cleaning process will improve heat transfer in the hydraulic fluid cooler, which can cause changes in the temperature and viscosity of the hydraulic fluid.

No. The plant technical specifications have been reviewed. It was found that the Technical Specifications are not affected by this tube cleaning operation.

SER NO.: 92-078 CROSS

REFERENCE:

DCP 91-3016A through E, Unit ¹2 DESCRIPTION OF CHANGE:

Modifyexisting pipe support/restraint configurations to reduce the total number of mechanical snubbers located on piping systems.

SUMMARY

No. As in the existing piping analysis, the applicable code design limits have been met to ensure piping integrity and system function; FSAR Sections 15.0.3.1, 15.0.3.5.

Also reviewed without impact were FSAR Sections 15.1, 15.2, 15.3, 15.4, 15.5, 15.6, 3.6.2, 15.7 and 15.8.

No. The proposed action only reduces the number of seismic and hydrodynamic restraints on selected piping systems by reanalyzing each line using optimum restraint configurations. The original allowable equipment interface loads are still being met and the pipe break criteria remains unchanged.

No. The margins of safety, as defined by the following Technical Specifications, will not be reduced: 3/4.4.1 Recirculation System, 3/4 4.9 RHR, 3/4.5.1 and 3/4.5.2 ECCS, 3/4.4.2 SRVs, 3/4.4.7 MSIVs, 3/4.7.3 RCIC, 3/4.3.1 RPS Instrumentation, 3/4.3.2 Isolation Actuation Instrumentation, 3/4.3.3 ECCS Actuation Instrumentation, 3/4.3.5 RCIC Actuation Instrumentation, 3/4.3.7 Monitoring Instrumentation, 3/4.4.8 Structural Integrity, 3/4.6.1 Primary Containment, 3/4.7.4 Snubbers.

SER NO.: 92-079 CROSS

REFERENCE:

DCP 91-9075, Unit ¹2 DESCRIPTION OF CHANGE:

Install two emergency lighting units in the vicinity of Panel 2C239 and four units along the Unit ¹2 Building north wall from Q to T line.

719'eactor

SUMMARY

No. The addition of emergency lighting units does not adversely affect safety-related systems or equipment. Electrical separation is maintained between the new units and the existing systems.

This modification does not interfere with the logic, control of operation of any safety-related plant system or components. This modification does not increase the probability of occurrence or the consequences of the accidents described in FSAR Chapters 6 and 15.

No. The proposed action does not affect the plant electrical system which provides normal and emergency AC power to reactor protection and engineered safety feature equipment. Failure of the emergency lighting units will not prevent the plant electrical system from performing its design safety functions. Accident scenarios have previously been analyzed in FSAR Chapters 6 and 15. This modification will not cause a different type of accident scenario other than those previously evaluated.

No. Technical Specification Basis 3/4.8.1, 3/4.8.2, and 3/4.8.3, "A.C. Sources, D.C.

Sources and Onsite Power Distribution" states, "The operability of the A.C. and D.C.

power sources and associated distribution systems during operation ensures that sufficient power will be available for (1) safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility." The proposed modification does not interfere with the logic, control, or operation of any safety-related plant system or component. The circuits feeding the individual emergency lighting units have breakers coordinated with upstream devices. Voltage drop and load ampacity are within design limits.

SER NO.: 92-080 CROSS

REFERENCE:

DCP 92-3013, Unit 42 DESCRIPTION OF CHANGE:

Installation of 3" manual ball valves in the condensate demineralizer vent, drain, and resin transfer line to provide a back-up means for positive isolation of the demineralizer vessels.

SUMMARY

No. FSAR Sections 10.4.6.1, "Condensate Cleanup System," 1.4.6.2, "External Regeneration System," and Section 15.7.2.1, "Miscellaneous Small Releases Outside Containment," have been reviewed. The additional ball valves are full port valves and are normally open during operations. The design integrity of the system is unaffected since the new manual valves fully comply with the original design codes.

The Condensate Demineralizer System is designed to function under normal operating conditions only. It has no safety-related function and does not directly interface with a safety-related system.

No. The installation of this modification will not affect the normal operation of the Condensate Demineralizer System. It merely provides redundant isolation to the condensate demineralizer vent lines and the resin transfer inlet and outlet lines for each demineralizer.

No. Technical Specification 3/4.11.1 establishes limits to unrestricted areas for liquid radwaste treatment system operability in minimizing radioactive material in liquid effluents prior to discharge. The addition of manual isolation valves in the condensate demineralizer vent lines and the resin transfer lines will not increase or add to these radioactive material limits.

SER NO.: 92-081 CROSS

REFERENCE:

DCP 91-9081A, 91-9081B, 91-9081Z, Unit ff2 DESCRIPTION OF CHANGE:

Modify LPCI inboard injection valves HV-252F015A8 B to alleviate pressure locking phenomena.

SUMMARY

No. This modification will be designed, constructed, and tested in accordance with the same code and design basis as the existing plant configuration. The valves being modified are ASME Section III, Class 1 valves and will be modified in accordance with a code repair plan per Section XI of the ASME Code.

Only a failure of both the valve packing and the upstream check valve to restrict flow in the reverse direction would create a significant release. In this case, the release through the valves would be limited by the 1/4" hole in the valve disc. This would constitute a small loss of coolant leak which is clearly bounded by the spectrum of line breaks evaluated in FSAR Section 6.3.

The radiological consequences of such a small leak are addressed in FSAR 15.7.2.

No. This modification does not add any active components; however, it does improve the reliability of these active valves. The installation of a pressure relief path between the bonnet cavity and the reactor side piping will eliminate a common mode failure mechanism identified in INPO 84-7, Bonnet Pressure Locking. No new failure modes will

~

be created. No new accidents or malfunctions have been created, No. Technical Specification requirements regarding the protection of the reactor coolant system boundary as they pertain to these valves are given in Section 3.4.3.2. In accordance with this section, leakage from any reactor coolant system pressure isolation valve is to be limited to one gallon per minute.

No change is being implemented which allows for system leakage to exceed the acceptance level approved by the NRC. Additional valve maintenance may be necessary; however, the margin of safety, i.e., leakage restricted to less than one gallon per minute, remains unchanged.

SER NO.: 92-082 CROSS

REFERENCE:

Operation of LLRWHF SSES, Common DESCRIPTION OF CHANCE:

The Low Level Radwaste Holding Facility (LLRWHF) storage plan and usage for the waste types and form generated at Susquehanna SES will be changed. Waste is expected to be stored for a period not to exceed five (5) years or until offsite disposal facilities are available.

SUMMARY

No. Handling and storage of waste containers is similar to previous actions. Fire potential within the LLRWHF has not increased with the proposed action. Probability for loss of offsite power remains the same with less radiological impact than previously analyzed.

Environmental accidents, i.e., freezes, floods, tornadoes and seismic events, and sabotage remain the same as previously analyzed.

The LLRWHF Operation proposed action shows an increase in dose rate and integrated dose for two accidents which are container dropped from transport vehicle and DAW fire.

The increase is below worst case accident previously evaluated in FSAR Chapter 15.

No. The LLRWHF is intended to be used for storage of low level radioactive waste and materials as previously designed and evaluated. High integrity containers, solidified-dewatered waste form, mobile shielded storage modules and crane(s) do not create a possibility for a different type of accident or malfunction.

No. The operation of the LLRWHF will be conducted in accordance with plant Technical Specifications. The following Technical Specifications were reviewed and it was concluded that there is no decrea'se in the margin of safety: 3/4.3.7.1, 3/4.3.7.9, 3/4.3.7.10, 3/4.3.7.11, 3/4.7.5, 3/4.7.6, 3/4.11.1, 3/4.11.2, 3/4.11.3, 3/4.11.4, 3/4.12.1, 5.1, 6.9.1.8, 6.12, 6.13, 6.14, and 6.15.

SER NO.: 92-083 CROSS

REFERENCE:

DCP 91-9083Z, 91-9083A, 91-9083B; Unit ff2 DESCRIPTION OF CHANGE:

This modification pertains to the Unit 42, System 251 Core Spray inboard injection valves. A 1/4" hole will be drilled through the valve disc face to the inner cavity behind the disc seats which connect the bonnet cavity to reduce the pressure.

SUMMARY

No. FSAR Section 5.2.5.3.2 acknowledges that certain components forming the reactor coolant boundary, such as valve packing, are not leak tight. Only a failure of both the valve packing and the upstream check valve to restrict flow in the reverse direction would create a significant release. The probability of a concurrent failure of both is less than that for similar leaks which can occur from systems carrying radioactive fluid.

No. The installation of a pressure relief path between the bonnet cavity and the reactor side piping will eliminate a common mode failure mechanism. This improves the reliability of these active valves. No new failure modes will be created.

No. Technical Specification Section 4.5.1 demonstrates these valves are operable by functional testing at least once per 18 months to verify that they actuate to the open position. This modification does not affect the ability of these valves to meet this requirement. In accordance with Technical Specification Section 3 4.3.2, leakage from any reactor coolant system pressure isolation valve is to be limited to one gallon per minute. This modification does not change the acceptance criteria in Technical Specification Section 3.6.1.2 for containment 'leakage.

SER NO.: 92-084 CROSS

REFERENCE:

DCP 92-3001, Units C1 8 02 DESCRIPTION OF CHANCE:

Installation of the necessary hardware and software for the existing plant process computer systems and a new ERDS Communication Personal Computer.

SUMMARY

No. The PCS and RDAS systems are not safety-related (FSAR Sections 7,5c and 7.7.1.7.1.2).

The modification of RDAS data-link parameters will not affect the POWERPLEX core monitoring system or STREAM effluent dose projection calculations. The PCS NSS SPOTMOS data-link is used to collect data from the plant SPOTMOS unit so that the Suppression Pool Bulk Temperature can also be monitored per Technical Specification surveillance requirements from the PCS's. The actual SPOTMOS unit is safety-related (FSAR Section 7.6.1b.1.2 4.2) and it will function without the NSS SPOTMOS data-link.

No. There are no safety-related systems in the EOF and the equipment does not interface with any safety-related equipment in the plant. The RDAS modifications do not change nor affect POWERPLEX core monitoring software calculations of safety-related parameters which are monitored per Technical Specification surveillance requirements.

No. The proposed hardware and software changes to RDAS, SPDS (EOF), ERDS PC and PCS

[data-link (software only)] do not affect the POWERPLEX fuel thermal margin calculations or Suppression Pool Bulk Temperature calculation.

SER NO.: 92-085 CROSS

REFERENCE:

SCP j921021; Unit ¹2 DESCRIPTION OF CHANGE:

Decrease the RBCCW heat exchanger discharge header low pressure alarm setpoint from 78 psig to 71 psig.

Implementation of this setpoint change will enhance operation of the RBCCW system by eliminating a nuisance alarm.

SUMMARY

No. The RBCCW system is discussed in FSAR Section 9.2.2. The affected pressure switch does not cause any automatic actions.

No. The device affected by the proposed setpoint change cannot affect the input conditions for accidents previously evaluated in FSAR.

No. Changing the setpoint of PSL-21308 will not affect the safety limits or limiting safety settings of equipment required for safety as described in the Technical Specifications.

SER NO.: 92-086 CROSS

REFERENCE:

OP-172-001 - Addition of Section 3.18, Units ¹1 5 ¹2 DESCRIPTION OF CHANGE:

Startup of the U-1 Reactor using the U-1 Recombiner supplying flow to U-1 'B'harcoal Subtrain and U-2 Subtrain. 'A'harcoal

SUMMARY

No. Section 15.7.1, Gaseous Radwaste Leak or Failure and specifically 15.7.1.1, Ambient Charcoal Offgas Treatment System Failure. Discussed are the failure of the system as a result of seismic, hydrogen explosions, fires in the assemblies and failure of spatially related equipment. The assumptions are that portion (or all) of the system is somehow damaged. The PCAF configuration, since it is a design feature of the system and involves only installed piping and components, is bounded by this analysis. Also, the system lineup does not increase the probability of these events occurring. The ability of the operator to address these failures and the response of the system does not change.

No. The proposed system lineup is a system design feature allowing the option to use absorber trains from either unit. The function of the absorber trains, as described in FSAR 11.3.2.2.2 is to remove moisture from the offgas flow prior to entering the absorber beds and to maintain releases with 10CFR20 limits. This function is not changed by transferring offgas from Unit ¹1 to Unit ¹2. In addition, the function of Unit ¹2 system is not affected since it will be using a separate isolated guard bed. The flow, as seen by the absorber bed, will be a combination of Unit ¹1 and Unit ¹2 offgas; but again, the function will not change. The effluent will be monitored and alarms will remain the same.

No. Technical Specification 3/4.11.2 addresses Gaseous Effluents. The proposed plant configuration will not alter the total offgas, and therefore, offsite does produced. The only difference is the distribution between the units. Unit ¹2 would have normal operating Offgas flow plus approximately one-half of Unit ¹1 Startup flow. This is still less than the flow that the absorber bed would see during a startup. The Technical Specification referenced provides various limits for maintaining offsite doses less than 10CFR Part 20.

These limits will remain the same, and this procedure will not place the plant in a configuration that will lead to exceeding these limits.

SER NO.: 92-087 CROSS

REFERENCE:

SCP CJ92-1018, AR-RW-007; Unit - Common DESCRIPTION OF CHANCE:

Change of setpoint of LRW Demineralizer Effluent Conductivity CSHL-06268 from 1.0/.8 ymhos/cm to 1.0/.1 ymhos/cm.

SUMMARY

No. FSAR Section 15.7.3 analyzes the complete release of the average radioactivity inventory in the tank containing the largest quantities of significant radionuclides in the liquid radwaste system, the evaporator concentrates waste tank. In addition, FSAR Section 2.4.13.3, Accident Effects, describes the potential effect on groundwater quality of an accidental release of liquid radwaste due to rupture of the same tank. This is a conservative water chemistry parameter change to the LRW demineralizer effluent conductivity instrumentation.

No. This does not result in any LRW equipment failure mode. FSAR Section 15.7.3 bounds any LRW system integrity failure.

No. This change does not affect any other radwaste processing setpoints.

SER NO.: 92-088 CROSS

REFERENCE:

Setpoint Change Packages for Generic Letter 89-10 MOV Torque Switch Settings; Units 1, 2 and Common DESCRIPTION OF CHANGE:

Issuing a setpoint change package that changes the Motor Operated Valve (MOV) allowable torque switch setting range as listed on the MOV Data Detail Drawing.

SUMMARY

No. Valve stroke time remains unaffected. Therefore, the valve stroke time still conforms to the stroke time requirements (as applicable as listed in FSAR Table 6.2.12).

The allowable torque switch setting range is developed in accordance with Design Standards MDS-04, MDS-03, MDS-01.

The motor electrical design criteria conform to FSAR 8.3.1.9.

The action does not adversely affect valve dynamic qualification. If the thrust range did not increase, the qualification basis remains unaffected. If the thrust range increases, an analysis is performed to assure that the valve assembly will be qualified at the specified acceleration levels determined from the piping analysis in conjunction with the higher thrust. Therefore, this action does not impact the dynamic qualification of the MOV as discussed in Section 3.9.3.2b.2 of the FSAR.

No. Does not adversely affect the ability of the MOV to perform its design intended function.

Does not affect the valve stroke time. Does not adversely affect the MOV dynamic qualification. And finally, does not affect any other equipment.

IIL No. AD-QA-423, "Station Pump Valve Testing Test Program," and AD-QA-412, "Leak Rate Test Program," referenced in the setpoint change package assures conformance to Technical Specification leakage requirements.

SER NO.: 92-089 CROSS

REFERENCE:

DCP ff91-9074; Unit ff1 DESCRIPTION OF CHANCE:

Installation of two emergency lighting units with corresponding remote heads in the vicinity of panel 1C239.

SUMMARY

No. The electrical coordination, separation requirements, diesel generator loading, voltage drop, circuit ampacity, combustible loading and safety impact design requirements and limits are met. FSAR Chapters 6 and 15 were reviewed.

No. The proposed action does not affect the plant electrical system which provides normal and emergency AC power to reactor protection and engineered safety feature equipment.

Failure of the emergency lighting units will not prevent the plant electrical system from performing its design safety functions. Accident scenarios have previously been evaluated in FSAR Chapters 6 and 15.

No. The proposed modification does not interfere with the logic, control, or operation of any safety-related plant system or component. The circuit feeding the individual emergency lighting units has a breaker coordinated with upstream devices. Voltage drop and load ampacity are within design limits. Technical Specification Basis 3/4.8.1, 3/4.8.2 and 3/4.8.3 were reviewed for this analysis.

SER NO.: 92-090 CROSS

REFERENCE:

DCP ff90-3036; Unit C2 DESCRIPTION OF CHANCE:

Replacing the existing Reactor Water Cleanup (RWCU) pumps with new sealless pumps; each with at least 100% system capacity.

SUMMARY

No. The following sections and tables of the FSAR were reviewed for applicability: Table 3.2-1 and 5.4-2; Sections 3.6.2.1.1.b) 7), 7.3.1.1a.2.4.1.9 and 10, 5.4.8.1, 2, and 3, 9.2.10 and 3.8.4.2.

The proposed modification to replace the existing pumps does not alter the function of the RWCU system.

No. The performance, function, classification and the flow capacity of the replacement pumps are the same as or better than that of the original pumps.

No. Technical Specification Sections 3.3.2, 4.3.2.1, 4.3.2.2 and 4.3.2.3 are applicable to RWCU. Table 3.3.2-2 lists the isolation actuation instrumentation trip setpoints for the system. Isolation actuation instrumentation and setpoints applicable to RWCU system are unchanged.

SER NO.: 92-091 CROSS

REFERENCE:

DCP ¹92-9011; Unit ¹2 DESCRIPTION OF CHANCE:

Connect spare conductor in parallel with conductors in existing cables and add a new cable to reduce voltage drop at 2C690A. Connect spare conductors in parallel with conductors in existing cables to reduce voltage drop at 2C690B.

SUMMARY

No. The proposed action enhances safety by reducing the voltage drops to the H,O, Analyzers during all plant conditions. This action assures that during degraded conditions, these analyzers receive sufficient voltage to operate. It also reduces the resistance of the circuit.

Chapters 6, 8, and 15 of the FSAR were reviewed.

No. The proposed action provides sufficient voltage at the analyzer terminals to assure operation. The existing design for the H,O, Analyzers has not been changed.

No. The proposed action does not reduce the margin of safety of the systems as defined in the basis of Technical Specification Sections: 3.3.7.5, 3.8.3.1 and 3.8.3.2.

SER NO.: 92-092 CROSS

REFERENCE:

DCP ¹91-9078; Unit ¹2 DESCRIPTION OF CHANGE:

This change will upgrade some of the pneumatic control components in the HVAC system to improve its reliability. It will reduce the alarm setpoint for high temperature in the MG Sets. It will add an alternate damper positioning capability.

SUMMARY

No. FSAR Sections 5.4.1, 7.4, 9.4.4, and 15.3.1 were reviewed for this evaluation. This modification will reduce the probability of tripping of the Reactor Recirculation Pumps due to loss of outside cooling air to the MG Sets by replacing the existing ventilation control components with improved materials for the tubing connectors and providing a properly selected range for differential pressure switch.

No. The only component failure which was not previously considered is the new pressure regulator. The effects of its failure or malfunction, however, in its role as backup following the failure or malfunction of the temperature controller, is no different than the failure of the controller itself. Use of the new alternate control mode replaces the present practice of bypasses or temporary repairs with an engineered system for use in the event of a normal control failure.

No. Technical Specification Basis 3/4.4.1, Reactor Recirculation System does not discuss the MG Sets or their ventilation and is unaffected by this modification.

SER NO.: 92-093 CROSS

REFERENCE:

DCP ¹92-9002; Unit Common DESCRIPTION OF CHANGE:

Repositioning of a solenoid operator and valve in the Standby Gas Treatment System. This was incorrectly installed and is being repositioned based on manufacturer's requirements.

SUMMARY

No. FSAR Chapter 15, Section 11.5 and 18.1.30 have been reviewed. No system function is altered and the design basis is not changed. This modification will enable the system to function as it was originally intended.

No. This modification will reroute the 3/4" diameter tubing and reinstall the existing valve in a vertical position. The valve will then operate as designed and the system will perform as intended.

No. The applicable Technical Specifications are Sections 3.3.7.5 and 3.3.7.11. The implementation of this modification will not adversely affect any system covered by the Technical Specifications, since the new and correct orientation of the solenoid will enhance the Post Accident Ventilation Stack Sampling (PAVSS) System's ability to monitor plant performance and provide information to operator as intended.

SER NO.: 92-094 CROSS

REFERENCE:

DCP ¹92-3013; Unit ¹2 DESCRIPTION OF CHANGE:

Installation of 3" manual ball valves in the vent, drain, and resin transfer lines of the Condensate Demineralizer System. This will provide a back-up means for positive isolation of the demineralizer vessels and provide the necessary blocking for maintenance of the air-operated butterfly valves.

SUMMARY

No. FSAR Sections 10.4.6.1, 10.4.6.2 and 15.7.2.1 have been reviewed. The additional ball valves are full port valves and are normally open during operations. The design integrity of the system is unaffected since the new manual valves fully comply with the original design codes. The Condensate Demineralizer System is designed to function under normal operating conditions only. It has no safety-related system.

No. This action provides redundant isolation to the condensate demineralizer vent lines and the resin transfer inlet and outlet lines for each demineralizer. The piping and supports in the areas where the new valves will be located will be qualified to original ANSI piping codes for the additional weight.

No. Technical Specification 3/4.11.1 establishes limits to unrestricted areas for liquid radwaste treatment system operability in minimizing radioactive material in liquid effluents prior to discharge. The addition to manual isolation valves in the condensate demineralizer vent lines and the resin transfer lines will not increase or add to these radioactive material limits. They will, in fact, help reduce the volume of LRW that must be processed by providing improved isolation during maintenance activities; thus, enhancing the system's performance and reliability.

SER NO.: 92-095 CROSS

REFERENCE:

DCP ¹92-9008; Unit ¹2 DESCRIPTION OF CHANC E:

Adding a time delay into the high system flow portion of the RWCU leakage detection actuation logic.

SUMMARY

No. A failure of the time delay relay or the PCIS actuation logic could prevent closure of the corresponding RWCU containment isolation valve. This single failure represents a random single failure for the division and is not a common mode failure since the circuitry for the PCIS actuation logic and RWCU containment isolation valve controls are designed to Class 1E criteria. Chapters 6 and 15 of the FSAR were reviewed for this evaluation.

No. The rewiring of the internal panel wiring in panels 2C622 and 2C623 is in accordance with existing approved installation termination procedures. The mounting of the time delays and the isolation relays was seismically analyzed. This resulted in the voltage at the devices and the loading of the buses and actuation control circuitry being acceptable.

No environmental qualification is required since the panels are located in the Control Structure which is a non-harsh environment. Electrical separation of the Class 1f and Non-Class 1E is in accordance with FSAR Section 8.1.6.1.q.

No. According to the following Technical Specifications that were reviewed, the margin of safety was not reduced: Section 3/4.6.3 and 3/4.3.2 and Tables 3.6.3-1 and 3.3.2-3.

SER NO.: 92-096 CROSS

REFERENCE:

DCP ¹91-9026; Unit ¹2 DESCRIPTION OF CHANCE:

The existing Rosemount Model 1151DP transmitters are replaced with Rosemount Model 1153 transmitters which are located in the RHR Service Water System. This replaces the obsolete transmitters with equivalent transmitters having a qualified life of 27 years.

SUMMARY

No. The FSAR was reviewed, in particular, Sections 6.7.4, 7.5 and 9.2.6, Chapter 15 and the responses to NRC questions. The new design does not introduce any new failure modes, so the overall system reliability is unchanged.

No. The new installation separates the square root extraction function from the transmitters.

This design change does not cause the loops to perform or function differently than the original loops.

No. Technical Specification 3/4.3.7.4 was reviewed. This change has no effect on the operability of the instrument loops.

SER NO.: 92-097 CROSS

REFERENCE:

CH-SY-015; Units ¹1 5 ¹2 DESCRIPTION OF CHANCE:

A non-oxidizing biocide and a clay slurry detoxicant will replace chlorine gas and sulfur dioxide detoxicant currently used for the control of biological growth in the circulating water and service water systems.

SUMMARY

No. Review of the various design basis accidents identified in Section 15 of the FSAR reveal that none of these accidents are affected by this treatment program since it does not impact the integrity, function, or performance of any plant safety-related equipment.

No. The Circulating Water System design basis is discussed in Section 10.4.5 of the FSAR. The change will reduce the corrosion of system components associated with microbiological control. There is no change to the design basis. There will be a positive impact on the operation of plant systems as a result of this change in treatment chemicals. The biocide concentration during treatments is negligible with respect to the total impurities in the circulating/service water. As such, it will have no measurable effect should a leak develop in closed cooling water heat exchangers or the main condenser.

No. The plant Technical Specifications have been reviewed and are not affected by this change.

SER NO.: 92-098 CROSS

REFERENCE:

DCP ¹91-9072; Unit ¹2 DESCRIPTION OF CHANCE:

The existing over and undervoltage relays located in 288,250 Volt DC power systems will be replaced with a C8 D Battery, Inc. relay. The intent is that with installation of the relays, the new setpoints established in calculation E-AAA-437, can be implemented.

SUMMARY

No. FSAR Table 8.3-22 evaluates the individual components that make up the 250 VDC system. Upon evaluation of the application of these relays and proposed setpoints, no increase in probability of an accident was found to occur. A failure of the relay to alarm on undervoltage would be picked up by the battery monitor on low battery voltage or battery voltage imbalance with an alarm input to the Control Room. A failure of the overvoltage relay is indicative of a battery charger problem or failure. The effects of a battery charger failure have been previously evaluated as part of FSAR Table 6.3-5.

No. The function of the new relays has not changed from that being replaced. The selectivity, repeatability and reset capability have been improved. The setpoints were chosen to provide early indication of system trouble.

No. The undervoltage setpoint actually increases the margin of safety by alerting the operator of a potential discharge of the battery while in its fully charged state. The 250 VDC will be out of service during implementation of this modification; thus, assuring personnel and plant safety.

SER NO.: 92-099 CROSS

REFERENCE:

DCP ¹91-9070; Unit ¹2 DESCRIPTION OF CHANGE:

The existing over and undervoltage relays located in 202,125 Volt DC power systems will be replaced with a C&D Battery, Inc. relay. The intent is that with installation of the relays, the new setpoints established in calculation E-AAA-437, can be implemented.

SUMMARY

No. FSAR Table 8.3-22 evaluates the individual components that make up the 250 VDC system. Upon evaluation of the application of these relays and proposed setpoints, no increase in probability of an accident was found to occur. A failure of the relay to alarm on undervoltage would be picked up by the battery monitor on low battery voltage or battery voltage imbalance with an alarm input to the Control Room. A failure of the overvoltage relay is indicative of a battery charger problem or failure. The effects of a battery charger failure have been previously evaluated as part of FSAR Table 6.3-5.

No. The function of the new relays has not changed from that being replaced. The selectivity, repeatability and reset capability have been improved. The setpoints were chosen to provide early indication of system trouble.

No. The undervoltage setpoint actually increases the margin of safety by alerting the operator of a potential discharge of the battery while in its fully charged state. The 250 VDC will be out of service during implementation of this modification; thus, assuring personnel and plant safety.

SER NO.: 92-100 CROSS

REFERENCE:

SCP j-91-1061; Unit ¹1 DESCRIPTION OF CHANGE:

Decreasing the setpoint of Cooling Tower Screen Differential Level Switches LDIS-11506 and LDIS-11507.

The effect of this change is to reduce the amount of screen blockage required to produce an alarm.

SUMMARY

No. Neither the switches, the screens, or the Circulating Water System are important to safety as evaluated in FSAR Section 10.4.5.1. The setpoint change does not affect any component or system. This implementation does not require any physical modifications. The decrease in differential level setpoint increases sensitivity to screen blockage which enhances, not degrades, the switches protective function.

No. The enhanced sensitivity allows more time for corrective action before the onset of pump cavitation. This proposed change ultimately will reduce the potential for loss of service water which would result in a reactor scram and thus reduces potential challenges to safety systems.

No. The switches are non-seismic, non-safety related, perform no function relative to safety, and their setpoint is not governed by Technical Specification. The screens and Circulating Water System also have no safety-related function.

SER NO.: 92-101 CROSS

REFERENCE:

SCP J-91-1019 and 1020; Units 01 8 t2 DESCRIPTION OF CHANGE:

Decreasing the setpoint of Cooling Tower Screen Differential Level Switches LDIS-11506 and LDIS-11507, LDISH-21506 and LDISH-21507. The effect of this change is to reduce the amount of screen blockage required to produce an alarm.

SUMMARY

No. Neither the switches, the screens, or the Circulating Water System are important to safety as evaluated in FSAR Section 10.4.5.1. The setpoint change does not affect any component or system. This implementation does not require any physical modifications. The decrease in differential level setpoint increases sensitivity to screen blockage which enhances, not degrades, the switches protective function.

No. The enhanced sensitivity allows more time for corrective action before the onset of pump cavitation. This proposed change ultimately will reduce the potential for loss of service water which would result in a reactor scram and thus reduces potential challenges to safety systems.

No. The switches are non-seismic, non-safety related, perform no function relative to safety, and their setpoint is not governed by Technical Specification. The screens and Circulating Water System also have no safety-related function.

SER NO.: 92-102 CROSS

REFERENCE:

SCP j-91-1062; Units ff1 5 02 DESCRIPTION OF CHANCE:

Increasing the setpoint of Cooling Tower Screen Differential Level Switches LDISH-21506 and LDISH-21507.

The effect of this change is to reduce the nuisance to the Control Room Operator, while still providing adequate sensitivity to blocked or clogged cooling tower screens.

SUMMARY

No. Neither the switches, the screens, or the Circulating Water System are important to safety as evaluated in FSAR Section 10.4.5.1. The setpoint change does not affect any component or system. This implementation does not require any physical modifications. The slight incr'ease in differential level setpoint eliminates an operating nuisance resulting from an oversensitive setpoint.

No. Other than the point of switch actuation, the change does not directly affect the performance of any component or system. The switch has an alarm only function and does not control any equipment.

No. The proposed change provides adequate time for the operator response following receiving the alarm, while still removing an operational nuisance.

SER NO.: 92-I03 CROSS

REFERENCE:

SCP J-91-1019 and 1020; Units ¹1 5 ¹2 DESCRIPTION OF CHANGE:

Decreasing the setpoint of Cooling Tower Screen Differential Level Switches LDIS-11506 and LDIS-11507, LDISH-21506 and LDISH-21507. The effect of this change is to reduce the amount of screen blockage required to produce an alarm.

SUMMARY

No. Neither the switches, the screens, or the Circulating Water System are important to safety as evaluated in FSAR Section 10.4.5.1. The setpoint change does not affect any component or system. This implementation does not require any physical modifications. The decrease in differential level setpoint increases sensitivity to screen blockage which enhances, not degrades, the switches protective function.

No. The enhanced sensitivity allows more time for corrective action before the onset of pump cavitation. This proposed change ultimately will reduce the potential for loss of service water which would result in a reactor scram and thus reduces potential challenges to safety systems.

No. The switches are non-seismic, non-safety related, perform no function relative to safety, and their setpoint is not governed by Technical Specification. The screens and Circulating Water System also have no safety-related function.

SER NO.: 92-104 CROSS

REFERENCE:

DCP 91-3016A through E; Unit ¹2 DESCRIPTION OF CHANGE:

The removal or replacement of a number of mechanical snubbers, as well as modifications to other pipe supports.

SUMMARY

No. The consequences of a LOCA event have been considered in FSAR Section 15.6.

Based on a thorough evaluation, it is reasonable to conclude that the probability of occurrence of a LOCA event will not increase due to the proposed modifications.

The following FSAR Sections were reviewed also: 15.0.3.1, 15.0.3.5 and 15.1.5.

No. The proposed action only reduces the number of seismic and hydrodynamic restraints (snubbers) on selected piping systems by reanalyzing each line using optimum restraint configurations and replaces two 2-inch Reactor Recirculation drain valves with functionally equivalent valves. FSAR Chapter 15 was reviewed.

No. A review of the following Technical Specification Sections for the proposed changes concluded that the safety functions of these systems would not be affected: 3/4.4.1, 3/4 4 9~ 3/4 5 1 3/4 5 2I 3/4 4 2 3/4 4 7~ 3/4 5 1 3/4 7 3I 3/4 3 1 3/4 3 2~ 3/4 3 3

~

3/4.3.5, 3/4.3.7, 3/4.5.1, 3/4.4.8, 3/4.6.1, 3/4.6.3 and 3/4.7.4.

SER NO.: 92-105 CROSS

REFERENCE:

Software Change; Units 41 8 /f2 DESCRIPTION OF CHANGE:

A change in the software used to solve the electrical load flow problem from ASDOP to CYME.

SUMMARY

No. The proposed change allows new software to be used in software quality level 2 applications. The net result is an improvement in the confidence level associated with the load flow solution results.

No. The application is calculational and is used to ensure that proper voltages are available at equipment buses. The software provides no operational function. It is off-line support software. Verification and validation have established the software's functionality.

No. The bases for Technical Specifications 3/4.8.1, 3/4.8.2 and 3/4.8.3 were reviewed.

The CYME software packages CYMBASE and CYMFLOW have been subjected to verification and validation activities which have demonstrated that CYMBASE and CYMFLOW function properly to yield a valid solution to the load flow problem. In application, this serves to ensure that necessary voltage is available at electrical buses.

SER NO.: 92-106 CROSS

REFERENCE:

DCP 91-3017, Unit ¹2 DESCRIPTION OF CHANCE:

Replacing the Pacific valves with new Anchor Darling parallel disc gate valves and new air-operated actuators.

The new valves will have a slotted yoke so that the yoke and cylinder subassembly may be slid onto the bonnet upper flange. This design feature will eliminate overhead interferences associated with having to raise the yoke and cylinder subassembly above the lower stem coupling.

SUMMARY

No. FSAR Sections 9.2.12.3, 3.2.1, 6.2.4 and Chapter 15 were reviewed for this evaluation. The new actuators will be environmentally and seismically qualified.

The Cv of the new Anchor Darling valves is virtually identical to that of the original Pacific valves which will have a negligible effect on the fluid drop. The new actuators are designed to open and close under normal and accident condition differential pressures. The piping stress levels have been reanalyzed to reflect the installation of the new valve and hanger modifications as necessary.

No. The possibility of a disc failure is created which cannot occur with the single piece disc in the existing Pacific valves. However, this type of failure is not considered realistic because the floating discs are designed with strict clearances and tolerances which will cause them to perform as well as the single piece discs. Minor problems with seat leakage have been experienced in other systems containing large amounts of crud. This is not expected to occur based on RBCW operational experience and due to the fact that the system uses demineralized water. FSAR Sections 3.2.1, 3.7A and 6.2.4 were reviewed.

III. No. Technical Specification Basis 3/4.6.1.1 ensures that the release of radioactive materials from containment is restricted to the leak paths and rates assumed in the accident analyses. Technical Specification Basis 3/4.6.1.2 ensures that the total containment leakage value will not exceed the value assumed in the accident analysis at peak accident pressure. Lastly, Technical Specification Basis 3/4.6.3 ensures that the containment atmosphere will be isolated from the outside environment in the event of radioactive release or containment pressurization.

A pressure locking evaluation was performed and it was determined that it could not prevent valves from performing their safety function. The wedge design of the parallel disc gate valve assures solid seating of the discs on both seating surfaces.

SER NO.: 92-107 CROSS

REFERENCE:

DCP 92-3017; Unit ¹2 DESCRIPTION OF CHANGE:

Re-wire the control circuits of Appendix R Safe Shutdown Path 2 Motor Operated Valves (MOVs). This modification is required for 10CFR, Appendix R compliance.

SUMMARY

No. The relocation of the limit and torque switches in the control circuit protects the Path 2 MOVs from an Appendix R fire in the Control Room. It ensures operability of the valves by enabling the action of limit and torque switches to control valve stroke under the postulated hot short condition.

No. The modification relocates the limit and torque switches "downstream" of the Control Room switches for the safe shutdown of Path 2 MOVs to ensure the operability of the valves in the event of an Appendix R fire. The design criteria for this modification is to maintain each MOVs functional requirements.

No. A review of the following Technical Specification Sections for the proposed changes concluded that the safety functions of these systems would not be affected: 3/4.3.3, 3/4.3.9, 3/4.5.1, 3/4.9.11, 3/4.3.5, 3/4.7.3, 3/4.3.4, 3/4.4.1, 3/4.4.4, 3/4.3.2 and 3/4.6.3.

SER NO.: 92-108 CROSS

REFERENCE:

DCP 92-9026; Unit ¹2 DESCRIPTION OF CHANGE:

Installation of two air louvers in each Branch Junction Module (BJM) enclosure to improve ventilation that will result in increased dependability for Reactor Manual Control System (RMCS) operation.

SUMMARY

No. The ventilation to the BJMs in accordance with vendor recommendations and instructions. The improved cooling accomplished by this modification will provide for increased dependability regarding RMCS operation and therefore, increased Unit

¹2 reliability. Failure of the louvers to perform properly would not prevent the RMCS equipment from performing its design function. FSAR Chapters 6 and 15 were reviewed.

No. Loss of the RMCS does not impair the reactor's ability to scram which ensures a safe condition for the plant.

No. This modification will only enhance the performance of certain equipment related to reactor control. The ability to insert/withdraw control rods as originally design will not be diminished. It will not impair the operation of any equipment or power supplies needed for safe shutdown or control of accident conditions. It does not adversely impact RMCS operability or surveillance criteria. Technical Specification 3/4.1.4.1, 3/4.1 and 3/4.1.4.2 were reviewed.

SER NO.: 92-109 CROSS

REFERENCE:

SCP ]92-1050; Unit ¹1 DESCRIPTION OF CHANGE:

Permanent adjustment of the high alarm setpoint of the Unit ¹1 CRD Rebuild Room Area Monitor, RI-13709, from 2.5 MR/Hr. to 25 MR/Hr. The purpose of this action is to more accurately reflect the historical radiological conditions during both outage and non-outage periods, more accurately reflect Health Physics postings and access control, provide a reasonable alarm level to alert personnel to both anticipated and unanticipated abnormally high radiation levels, and minimize the potential for "annunciator masking."

SUMMARY

No. The subject ARM provides only alarm and indication functions. It is also not safety-related. FSAR Sections 12.3.4.1.5, 12.3.4.1, 12.5 and Table 12.3-7 were referenced.

No. During CRD repair activities, the CRD Rebuild Room will become a higher dose rate area which was estimated to be 15 MR/Hr. The general area radiation levels in the Unit ¹1 CRD Rebuild Room are above the existing ARM setpoint of 2.5 MR/Hr. The FSAR clearly recognizes that it is reasonable and prudent to adjust any ARM based operational considerations when supported by measured radiation levels. FSAR Section 12.3 and Table 12.3-7 were reviewed.

No. The subject ARM is not addressed in the bases for any Technical Specifications.

SER NO.: 92-110 CROSS

REFERENCE:

SCP ]92-1051; Unit ¹1 DESCRIPTION OF CHANGE:

Permanent adjustment of the high alarm setpoint of the Unit ¹1 CRD Rebuild Room Area Monitor, RI-1 3709, from 2.5 MR/Hr. to 25 MR/Hr. The purpose of this action is to more accurately reflect the historical radiological conditions during both outage and non-outage periods, more accurately reflect Health Physics postings and access control, provide a reasonable alarm level to alert personnel to both anticipated and unanticipated abnormally high radiation levels, and minimize the potential for "annunciator masking."

SUMMARY

No. The subject ARM provides only alarm and indication functions. It is also not safety-related. FSAR Sections 12.3.4.1.5, 12.3.4.1, 12.5 and Table 12.3-7 were referenced.

No. During CRD repair activities, the CRD Rebuild Room will become a higher dose rate area which was estimated to be 15 MR/Hr. The general area radiation levels in the Unit ¹1 CRD Rebuild Room are above the existing ARM setpoint of 2.5 MR/Hr. The FSAR clearly recognizes that it is reasonable and prudent to adjust any ARM based operational considerations when supported by measured radiation levels. FSAR Section 12.3 and Table 12.3-7 were reviewed.

No. The subject ARM is not addressed in the bases for any Technical Specifications.

SER NO.: 92-111 CROSS

REFERENCE:

"DCP 92-9006; Unit ¹2 DESCRIPTION OF CHANGE:

Change the cleanup line isolation signal from reactor vessel low water level 3 (+13") or high drywell pressure to level 2 (-38") or high drywell pressure. This increases the margin between water levels during power operation and the isolation level; and therefore, increases the likelihood that the cleanup line will be available following transients.

SUMMARY

The isolation signal change from level 3 to 2 does not degrade the operation of any equipment. The Agastat relays and level switches to be used for the level 2 isolation signals are identical in design, material, and construction to those currently used for level 3. FSAR Sections 6.2, 6.3 and Chapter 15 were reviewed.

The modification uses spare terminals on existing relays to receive an input from a level 2 switch. These relays are identical to those used to receive the level 3 input signal. The actuation logic remains single failure proof. The change in internal wiring is in accordance with existing installation and termination procedures. The use of the cleanup line serves to lower the pool water level which decreases the loading during a LOCA or safety/relief valve operation. This is in accordance with FSAR Sections 6.2, 6.3, and Chapter 15.

No. The reactor vessel low water level signal for isolation of the suppression pool cleanup is contained in Technical Specification Table 3.6.3-1. It indicates level 3 for Unit ¹2. This is not explicitly stated in the bases of Technical Specification.

SER NO.: 92-112 CROSS

REFERENCE:

DCP 92-9019Z; Unit ¹2 DESCRIPTION OF CHANGE:

Add redundancy to each Unit ¹2, 4.16 KV bus degraded voltage timer reselect logic. The redundancy prevents a random single failure from affecting the operability of more than one channel of safety-related equipment.

SUMMARY

No. The change does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relays in plant auxiliary load shed scheme and the additional contacts in the degraded voltage timer reselect logic.

However, based on engineering judgement, the increase in probability is considered to be small or insignificant enough so that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of overall system malfunction. Chapters 6 and 15 of the FSAR were reviewed.

No. The wiring and terminations of the external field cables and internal panel wiring within panels 2C221 and 2C222 is in accordance with existing approved installation termination procedures. The mounting of the isolation relays was seismically analyzed. Since the panels 2C221 and 2C222 are located in the Reactor Buildin'g, Secondary Containment, the relays are environmentally qualified for harsh environment. A random single failure during a LOCA/LOOP has the same impact as a malfunction with the existing Plant Auxiliary Load Shed and the degraded Voltage Reselect Logic. FSAR Chapters 6 and 15 were reviewed.

m. No. The proposed action does not affect the required number of operable channels, the setpoints or the response times. This evaluation was based on review of Technical Specification Section 3/4.3.3 and Tables 3.3.3-1, 2, 3.

SER NO.: 92-113 CROSS

REFERENCE:

TP-069-040; Unit - Common DESCRIPTION OF CHANGE:

LRW Collection Tank Organic Reduction using Hydrogen Peroxide.

SUMMARY

No. The proposed action involves injection of test chemicals to the LRW collection system and subsequent processing through the LRW filters and demineralizer to the sample tanks. The proposed test does not increase the radioactive inventory in any radwaste tank and does not involve the evaporator processing stream including the evaporator concentrates waste tank and associated piping and support equipment.

FSAR Sections 15.7.3 and 2.4.13.3 were reviewed for this analysis.

No. Since the test does not affect the radioactivity contained in any LRW system, a failure as a result of the test would be no different than any possible failure. FSAR Section 15.7.3, involving failure of the evaporator concentrates tank, bounds all LRW integrity failures.

No. The proposed LRW processing change is intended to remove organic material from the liquid radwaste stream and does not involve (or change) the radioactivity that is stored or transferred in any waste stream. Technical Specification Sections 3/4.3.7, 3/4.4.4, 3/4.11, 3/4.12 were reviewed.

SER NO.: 92-114 CROSS

REFERENCE:

SCP-]92-1048; Unit ff1 DESCRIPTION OF CHANGE:

Increase of isolation setpoints to provide margin comparable to the Reactor Building steam tunnel and to ensure that non-leak conditions will not cause a main steam line isolation.

SUMMARY

No. The radiological consequences of a 65 gpm leak fall well below the Standard Review Plan acceptance criteria and below the dose for the steam line break analysis in the FSAR. Increasing the temperature switch setpoints allows for higher leakage without automatically isolating the main steam lines. FSAR Sections 15.6, 7.3.1.1a.2.4.1.3, 15.24 and Table 15.6-9 were reviewed.

No. Through wall leakage is a precursor condition for a steam line break accident analyzed in FSAR Section 15.6.4.

No. The temperature switches and setpoints are listed in Technical Specification Section 3.3.2. These changes support this basis.

SER NO.: 92-115 CROSS

REFERENCE:

SCP-j92-1049; Unit ¹2 DESCRIPTION OF CHANCE:

V Increase of isolation setpoints to provide margin comparable to the Reactor Building steam tunnel and to ensure that non-leak conditions will not cause a main steam line isolation.

SUMMARY

No. The radiological consequences of a 65 gpm leak fall well below the Standard Review Plan acceptance criteria and below the dose for the steam line break analysis in the FSAR. Increasing the temperature switch setpoints allows for higher leakage without automatically isolating the main steam lines. FSAR Sections 15.6, 7.3.1.1a.2.4.1.3, 15.24 and Table 15.6-9 were reviewed.

No. Through wall leakage is a precursor condition for a steam line break accident analyzed in FSAR Section 15.6.4.

No. The temperature switches and setpoints are listed in Technical Specification Section 3.3.2. These changes support this basis.

SER NO.: 92-116 CROSS

REFERENCE:

DCP 92-9032; Unit ¹2 DESCRIPTION OF CHANCE:

Replacement of the 2X101A main transformer, rework of the existing fire protection piping system around the 2X101A transformer to provide spray coverage, installation of new transformer neutral to ground connection, modify MCCs 28101081, 2B111041 by removing ground relays, CTs and wiring, rework conduit 2KA065 at the 2X101A control cabinet and install new terminal blocks.

SUMMARY

No. The replacement transformer is functionally compatible with the existing McGraw-Edison transformer in its ability to provide required voltage and current to the existing 500 KV transmission network. The other modification work is done to supplement the transformer replacement so as to satisfy existing codes/requirements and to provide the proper secondary equipment needed for satisfactory operation. FSAR Chapters 6 and 15 were reviewed.

No. Failure of the 2X101A main transformer unit will not prevent the plant electrical system from performing its design safety functions. Accident scenarios have previously been analyzed in FSAR Chapters 6 and 15.

No. The proposed modification does not interfere with the logic, control, or operation of any safety-related plant system or component. The performance characteristics between the existing McGraw-Edison and the replacement Westinghouse transformer are essentially identical as delineated by Technical Specification E-1057. Technical Specification Basis 3/4.8.1, 3/4.8.2 and 3/4.8.3 were reviewed.

SER NO.: 92-117 CROSS

REFERENCE:

DCP 90-3097; Unit ff2 DESCRIPTION OF CHANGE:

Replace the existing Pacific wedge gate valves which have a long history of failure to open due to thermal binding. The replacement valves will be parallel slide gate valves which are less susceptible to thermal binding and are designed to operate in systems which experience thermal transients.

SUMMARY

No. The valves and associated piping, where applicable, are being welded into the existing system using approved procedures in order to minimize the chance of leakage. The valves are located in the turbine building and no safety equipment is installed below them. The existing air cylinder lubrication oilers will be removed which will actually reduce the fire loading in this area. This evaluation is in accordance with FSAR Sections 15A.6.3 and 10.4.10.3.

No. The replacement valves will perform the same function as the existing valves. The valves have been leak tested by the vendor to assure minimal seat bypass leakage.

No. The new valves and piping do not change the function or operation of the system in which they are to be installed.

SER NO.: 92-118 CROSS

REFERENCE:

DCP 89-3013C; Unit ¹2 DESCRIPTION OF CHANCE:

Installation of a second isolation valve upstream of the existing vent or drain line isolation globe valve and the removal of the existing globe valve by cutting and capping the pipe. This modification will significantly reduce/completely eliminate leakage of feedwater into the radwaste system, and enhance system availability.

SUMMARY

No. FSAR Sections 10.4.7, 10.4.10 and Chapter 15 were reviewed for this analysis.

During operation, radioactive steam is present in the extraction steam piping and the feedwater heater shells. The valves being replaced are installed in piping which connects to these barriers. FSAR Section 15.6.4 addresses steam system piping breaks outside containment. The building fault event for breaks outside containment is the complete severance of one of the four main steam lines. The calculated exposure for this accident is illustrated in FSAR Table 15.6-9 and represents only a small fraction of the 10CFR100 guidelines. This scenario envelops the same failure for the pipe being modified.

This modification does not jeopardize the function or alter the operation of any safety-related equipment.

No. This modification is non-Q and only interfaces with non-Q systems. In addition, the installation and design will be in accordance with all original codes and standards.

The replacement valves and piping, where applicable, will be welded into the existing system as are the original valves and piping in order to minimize the risk of leakage in accordance with FSAR Section 10.4.10.3. All components being replaced or added are of similar physical characteristics to the original components.

No. The only Technical Basis that discusses the feedwater system is 3/4.3.9 which is not affected by the proposed action. This modification is non-safety related and does not affect systems having Technical Specification requirements. This change adds valves which are designed and built to the same standards and design conditions as the feedwater heater drain and vent system. The addition of new valves or the removal of existing valves by cutting and capping the pipe will not in any way change the operation of the feedwater heater drain and vent system.

SER NO.: 92-119 CROSS

REFERENCE:

DCP 92-9039Z; Unit ¹2 DESCRIPTION OF CHANCE:

Add a redundant LOCA start signal to the Diesel Generator D BACKUP Auto Start logic. The redundancy prevents a random single failure of the Unit ¹2 Channel B Battery from affecting the operability of more than one channel of safety-related equipment.

SUMMARY

No. The addition of a redundant LOCA Start signal into the Diesel Generator D BACKUP Auto Start logic does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relay in the Division I RHR LOCA logic and the additional contacts of the relay and test switch in the Diesel Generator D BACKUP Auto Start logic. However, based on engineering judgement, the increase in probability is considered to be so small or insignificant so that the change is within the error bounds associated with the original design calculations and does not constitute a significa'nt increase in probability of overall system malfunction.

No. The wiring and terminations of the internal panel wiring within panels 2C617 and 2C618 are in accordance with existing approved installation termination procedures.

If a random single failure occurred during normal operation, the impact is the same impact as a malfunction with existing BACKUP Auto Start logic. FSAR Chapters 6 and 15 and Section 8.1.6.l.q were reviewed.

No. The proposed action does not affect the operability requirements for AC sources (Section 3/4.8.1) or ECCS Instrumentation (Section 3/4.3.3) including Tables 3.3.3-1, 3.3.3-2 and 3.3.3-3.

SER NO.: 92-120 CROSS

REFERENCE:

DCP 92-9037Z; Unit ¹2 DESCRIPTION OF CHANCE:

Add a redundant LOCA start signal to the Diesel Generator D BACKUP Auto Start logic. The redundancy prevents a random single failure of the Unit ¹2 Channel A Battery from affecting the operability of more than one channel of safety-rela'ted equipment.

SUMMARY

No. The addition of a redundant LOCA Start signal into the Diesel Generator D BACKUP A'uto Start logic does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relay in the Division I RHR LOCA logic and the additional contacts of the relay and test switch in the Diesel Generator D BACKUP Auto Start logic. However, based on engineering judgement, the increase in probability is considered to be so small or insignificant so that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of overall system malfunction.

No. The wiring and terminations of the internal panel wiring within panels 2C617 and 2C618 are in accordance with existing approved installation termination procedures.

If a random single failure occurred during normal operation, the impact is the same" impact as a malfunction with existing BACKUP Auto Start logic. FSAR Chapters 6 and 15 and Section 8.1.6.1.q were reviewed.

m. No. The proposed action does not affect the operability requirements for AC sources (Section 3/4.8.1) or ECCS Instrumentation (Section 3/4.3.3) including Tables 3.3.3-1, 3.3.3-2 and 3.3.3-3.

SER NO.: 92-121 CROSS

REFERENCE:

DCP 92-9017Z; Unit ¹2 DESCRIPTION OF CHANCE:

A relay will be added to the HPCI F006 injection valve open indication circuit to provide a new permissive for the transfer.

SUMMARY

No. The new relay cannot cause an inadvertent HPCI start up as discussed in FSAR Section 15.5.1. The probability of a failure of the new relay has been minimized by use of a Class 1E seismically and environmentally qualified relay of the same type already employed in the ESF control systems. The operability of the new relay will be confirmed by periodic functional testing. The increase in failure rate due to the change falls within the error band of the predicted failure rate of the valves without the additional relay. This modification does not alter the ECCS response during accident operating modes involving HPCI injection. FSAR Sections 7.3, 6.3, 15.2.4 and 6.2 were reviewed for this evaluation.

No. The malfunction of the new relay could result in the loss of control of the injection valve F006 or the loss of the automatic pump suction transfer function on high suppression pool water level. Failure of the injection valve to open would constitute loss of the HPCI ECCS function. This accident is discussed in FSAR Section 6.3. The automatic transfer on high suppression pool water level is only one mode of initiating transfer, and its loss is bounded by the potential loss of function of the F042 suction valve itself. This also results in a loss of HPCI ECCS function for which the ADS system serves as backup as described in FSAR Section 6.3.

III. No. Failure of the new relay to operate will preclude automatic pump suction transfer on high suppression pool level during HPCI injection. However, the existing design of the automatic transfer is already subject to a number of single failures which could result in the failure of the F042 valve to transfer the HPCI pump suction.

The maximum permissible suppression pool water level is 24'-0" which is in accordance with Technical Specification 3.6.2.1.

SER NO.: 92-122 CROSS

REFERENCE:

DCP 90-3055F, Rev. 1; Unit 02 DESCRIPTION OF CHANCE:

The existing air-operated butterfly valves will be replaced with the new control valves having equal percentage control characteristics, high turndown ratio and increased anti-activation protection. This will improve the chiller condenser recirculation loop flow control to maintain the condenser service water outlet temperature at 105'F.

SUMMARY

No. As discussed in FSAR Section 9.2.12.3, the only safety-related functions of the RBCW System are performed by the primary containment piping penetrations and containment isolation valves. As stated in FSAR Section 9.2.1, the SW System is designed to operate during normal plant operation and plant shutdown with offsite power available. The SW System will not operate on loss of offsite power concurrent with a LOCA.

No. The new temperature controller will maintain constant chiller condenser cooling water outlet temperature by modulating temperature control valves on the cooling water return and recirculation lines. FSAR Sections 9.2.1.1, 9.2.1.2, 9.2.12.3 and 9.2.12.3.1 were reviewed for this evaluation.

No. The RBCW System and the SW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for average drywell temperatures during reactor operating conditions 1, 2 and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

SER NO.: 92-123 CROSS

REFERENCE:

DCP 90-3060F, Rev. 1; Unit 02 DESCRIPTION OF CHANGE:

The existing air-operated butterfly valves will be replaced with the new control valves having equal percentage control characteristics, high turndown ratio and increased anti-activation protection. This will improve the chiller condenser recirculation loop flow control to maintain the condenser service water outlet temperature at 105~F.

SUMMARY

No. As discussed in FSAR Section 9.2.12.3, the only safety-related functions of the RBCW System are performed by the primary containment piping penetrations and containment isolation valves. As stated in FSAR Section 9.2.1, the SW System is designed to operate during normal plant operation and plant shutdown with offsite power available. The SW System will not operate on loss of offsite power concurrent with a LOCA.

No. The new temperature controller will maintain constant chiller condenser cooling water outlet temperature by modulating temperature control valves on the cooling water return and recirculation lines. FSAR Sections 9.2.1.1, 9.2.1.2, 9.2.12.3 and 9.2.12.3.1 were reviewed for this evaluation.

No. The RBCW System and the SW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for average drywell temperatures during reactor operating conditions 1, 2 and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

SER NO.: 92-124 CROSS

REFERENCE:

DCP 92-3015; Unit Common DESCRIPTION OF CHANGE:

Addition of sample tags and corrosion probes to the D/G closed loop jacket water piping.

SUMMARY

No. The SAR has been reviewed; specifically Chapter 15, "Accident Analysis," and Chapter 9, Section 9.5.5, "Diesel Generator Cooling Water System."

These attachments to the piping will provide a means to sample, analyze and maintain the proper chemical composition in the water to minimize corrosion and prevent microbiological activity. These sampling and monitoring points do not alter the Jacket Water System's function and the design bases are not changed. The system will function normally with the increased ability to sample and monitor corrosion rates.

No. The corrosion probe electrodes are engineered to withstand the flow of the system and the likelihood of an electrode detaching from the probe is remote. However, if an electrode would get loose in the system, the size is small enough that it would pass through the system ending up in the standpipe with no affect on system components or system operability.

Providing the jacket water system with monitoring capabilities does not adversely affect the operation of the system. It will, however, enhance the ability to monitor and control the corrosion rate within the system and determine the exact amount of chemicals the system requires.

I II. No. The Technical Specifications have been reviewed, specifically Section 3/4.8, "Electrical Power Systems." Since the original piping codes and standards are utilized, the implementation of this modification will not change the function or affect the operation of any component or system related to the diesel generators. The addition of the corrosion probes and sample taps simply allows for better system maintenance by being able to properly regulate the amount of corrosion inhibitors added to the system.

SER NO.: 92-125 CROSS

REFERENCE:

DCP 92-9042; Unit ¹2 DESCRIPTION OF CHANGE:

Replace the existing Reactor Feed Pump Turbine (RFPT) Steam Admission Valves HV-22710A, B and C with new valves to increase system performance and improve the environmental conditions around the valves.

SUMMARY

No. A review of the following FSAR Sections for the proposed changes concluded that there is no increase in probability of occurrence or the consequences or malfunction of equipment important to safety: 10.3, 10.4.7, 15.0, 1.2.2.8.15, 5.2.5.2, 5.4.12 and 11.3.2.4.3.

No. Replacing these valves with a better valve will have no adverse affects on any system or system functions and will increase system performance and improve the environmental conditions of the area where the valves are located. The removal of the tubing on the Process Valve Stem Leakoff Collection System will provide a more open area to work around the turbine and a potential condenser leak will be eliminated with the removal of the leakoff isolation valves.

No. The Unit ¹2 Technical Specifications have been reviewed, specifically Section 3/4.11.2, "Gaseous Effluents." Replacing the steam admission valves will have no adverse effects on the Main Steam Supply System or the RFPT; it will, however, enhance the system's ability to operate with the addition of more reliable valves.

The removal of the Process Valve Steam Leakoff Collection System from these valves will have no affect on the offsite iodine release rate limits given in Section 3/4.11.2 of the Unit ¹2 Technical Specification.

SER NO.: 92-126 CROSS

REFERENCE:

DCP 92-9062A, B, C, D; Unit Common DESCRIPTION OF CHANCE:

Installation of vents in the 1/4" fuel oil drain lines which run from each of the diesel generator cylinder heads to the respective fuel injection pump's pedestal chamber to provide better flow.

SUMMARY

No. The SAR has been reviewed; specifically Chapter 15, "Accident Analysis" and Chapter 9.5, "Other Auxiliary Systems."

This modification deals with adding vendor recommended vents in the 1/4" fuel oil drain tubing from each of the 16 cylinder heads to its respective fuel injection pump.

The drain system which runs to the fuel oil day tank is a system that is designed to be vented. The header pipe which collects the fuel oil from each of the injector pumps is vented along with the fuel oil storage tank. This modification will only add additional venting to the system upstream of the injector pump. The function of the Diesel Generators will not be degraded but, in fact, enhanced by this modification.

No. The new vent lines added will conform to the same requirements, both for installation and service, as the existing fuel oil tubing. The actual operation of the fuel oil drain system will be improved by providing better flow through the injection pump pedestal. The modification is an enhancement recommended by Cooper-Bessemer.

No. The Technical Specifications have been reviewed, specifically Section 3/4.8, "Electrical Power Systems." Since the original piping codes and standards are utilized, the implementation of this modification will not change the function or adversely effect the operation of any component or system related to the emergency diesel generators. The addition of these vents to providing better flow through the injection pump will only improve the system reliability and ability to operate as intended.

SER NO.: 92-127 CROSS

REFERENCE:

DCP 92-3012; Unit ¹2 DESCRIPTION OF CHANGE:

Replacement of the worm, worm gear, motor pinion and worm shaft gear on HV-255F022. This will increase the maximum torque switch setting which allows the field to increase the actual torque switch setting, thereby improving the probability of obtaining acceptable results in future static diagnostic testing. This is performed in the HPCI system.

SUMMARY

No. After implementation of the modification, the MOV isolation time will be increased.

This does not present a concern since the new design isolation time is 34.8 seconds which is well within the FSAR and Technical Specification limit of 50 seconds.

FSAR Sections 15.6, 3.9.3.2b.2 and 6.3.2.2.1 were reviewed.

No. The change does not adversely impact the ability of the MOV to isolate against conditions associated with high energy line break. It does not affect the ability of HPCI to perform its design intended function as described in Section 6.3.2.2.1 of the FSAR. It does not increase the stroke time beyond that previously contained in the Design Basis for SSES (FSAR Table 6.2-12). It does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis. Finally, it does not affect any other equipment.

No. The bases for Technical Specifications 3/4.3.2 and 3/4.3.3 discuss reactor system isolation actuation instrumentation and ECCS system actuation instrumentation. Since the modification affects no equipment other than HV-255F002, these bases are unaffected. The modification has no effect on HPCI operation and therefore does not necessitate a change to the basis for Technical Specification 3/4.5, "Emergency Core Cooling Systems." With regard to the containment isolation function of HV-255F002, the integrity of the valve will not be adversely affected and the isolation time remains within the accident analysis bounds of 50 seconds; therefore, the bases for Technical Specification 3/4.6.1, "Primary Containment," and Technical Specification 3/4.6.3, "Primary Containment Isolation Valves," are unaffected.

SER NO.: 92-128 CROSS

REFERENCE:

DCP 92-3021; Unit ¹2 DESCRIPTION OF CHANCE:

Replacement of the worm, worm gear, motor pinion and worm shaft gear on HV-'251F004A. This will satisfy the requirements of both the reduced voltage pull out and the LLRT. This is performed in the RHR system.

SUMMARY

No. After implementation of the modification, the MOV time will be increased from 123 seconds to 200 seconds. This has no adverse affect on the valve's safety function because it does not receive any automatic isolation signals. FSAR Sections 15.6 and 3.9.3.2b.2 were reviewed for this analysis.

No. This modification does not affect the ability of RHR to perform its design intended function as described in FSAR Sections 5.4.7 and 6.3.2.2.4. It does not affect any system requirements for a specific stroke time. It does not alter the MOV's allowable seismic acceleration. It does not affect any other equipment.

No. The bases for Technical Specifications 3/4.3.2 and 3/4.3.3 discuss reactor system isolation actuation instrumentation and fCCS system actuation instrumentation. Since the modification affects no equipment other than HV-251F004A, these bases are unaffected. The modification has no effect on LPCI operation since the valve is normally open and therefore does not necessitate a change to the bases for Technical Specification 3/4.5, "Emergency Core Cooling Systems." Alignment of the RHR system from LPCI to shutdown cooling and back is performed as a manual operation.

The modification has no effect on shutdown cooling operation or realignment to LPCI since the valve cycle is manually initiated, and therefore does not necessitate a change to the bases for Technical Specification 3/4.4.9, "Residual Heat Removal."

With regard to the containment isolation function of HV-251F004A, the integrity of the valve is not adversely affected and it remains under remote manual control without a specified closure time per Table 3.6.3-1; therefore, the bases for Technical Specification 3/4.6.1, "Primary Containment," and Technical Specification 3/4.6.3, "Primary Containment Isolation Valves" are unaffected.

SER NO.: 92-129 CROSS

REFERENCE:

DCP 92-5001/92-5002; Units ¹1 5 ¹2 DESCRIPTION OF CHANCE:

Installation of a timer circuit to monitor the data gathering functions of the Sentinel Program.

SUMMARY

No. The GETARS computer is not safety-related per FSAR 7.7.1.9. Installation of the proposed modification will enhance the operator's awareness of the GETARS availability. This change does not increase the probability or consequences of an accident since GETARS computer does not interface with any safety-related systems.

The GETARS computer is not essential for safe shutdown of the plant and serves no active emergency function during operations.

No. The GETARS computer is not safety-related per FSAR 7.7.1.9. The proposed modification does not change design criteria for any safety-related system or function as described in the FSAR.

No. The GETARS computer is not addressed in the Technical Specifications. The SSES Technical Sections B2 and B3/4 were reviewed and no conflicts exist. The intent of the modification is to improve the operability of the GETARS computer and thereby increase operator confidence.

SER NO.: 92-130 CROSS

REFERENCE:

DCP 92-9029Z, Unit ¹2 DESCRIPTION OF CHANCE:

Reconnect the Unit ¹2 LOCA interlock in the Unit ¹1 Plant Auxiliary Load Shedding Initiation circuit so that the Unit ¹1 degraded voltage timers reselect from 5 minutes to 9 seconds time delay on a Unit ¹2 LOCA.

SUMMARY

No. The modification enhances the function of the circuit in responding to accident conditions. It does not affect any of the postulated events identified in Chapters 6 and 15 of the FSAR. Based on engineering judgement, any increase in the probability of an equipment malfunction as a result of the proposal action is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of overall system malfunction. The slight increase is offset by the improvement in the response of the load shed logic to a Unit ¹2 LOCA.

No. A random single failure in either logic during normal operation has no impact since the circuitry for both logics is deenergized. A random single failure during a LOCA/LOOP has the same impact as a malfunction with the existing Plant Auxiliary Load Shed Logic.

No. The operability of the Degraded Voltage Timer Reselect Logic is governed by Technical Specification Section 3/4.3.3 for each unit, entitled, "Emergency Core Cooling System Actuation Instrumentation," with Tables 3.3.3-1, 3.3.3-2, and 3.3.3-3 establishing the required number of operable channels, setpoints and response times.

The bases for operability of the Degraded Voltage Timer Reselect Logic is to ensure that the Emergency Core Cooling System Actuation Instrumentation can provide the initiating actions to mitigate the consequences of accidents that are beyond the ability of the operator to control.

SER NO.: 92-131 CROSS

REFERENCE:

DCP 92-9041Z; Unit ¹2 DESCRIPTION OF CHANCE:

1. Replace the 93 percent Degraded Voltage Relays (ITE Type 27D) with similar degraded voltage relays (ABB Type 27N) having a narrower, adjustable dead band.

2. Change the time delay setpoint of the 93 percent Degraded Voltage Relays from 1 second to 7 seconds.

3. Change the time delay setpoint of the Degraded Voltage LOCA Timer Relays from 9 seconds to 3 seconds.

4. Change the time delay setpoint of the Degraded Voltage Alarm Relays from 9 seconds to 3 seconds.

This modification is done to prevent an initiation of a degraded voltage protection scheme operation that causes a delayed Loss of Offsite Power (LOOP).

SUMMARY

No. The ABB Type 27B Degraded Voltage Relay is from the same product line as the original ITE Type 27D Relay. The Type 27N Relay is an evolution of the Type 27D Relay. The reliability of the Type 27N is comparable to the Type 27D. The time delay setpoint change does not affect the reliability of the relays. FSAR Chapters 15 and 6 were reviewed.

No. The determinating and reterminating of existing internal panel wiring for the relays is in accordance with existing approved installation termination procedures. A random single failure in the Degraded Voltage Protection logic does not create a malfunction of a different type. Chapters 6 and 15 of the FSAR were reviewed.

No. The operability of the Degraded Voltage Relays and the Degraded Voltage Timer Relays is governed by Technical Specification Section 3/4.3.3, entitled, "Emergency Core Cooling System Actuation Instrumentation," with Tables 3.3.3-1, 3.3.3-2, and 3.3.3-3 establishing the required number of operable channels, the setpoints and response times. The bases for operability of the Degraded Voltage Relays and the Degraded Voltage Timer Relays is to ensure that the Emergency Core Cooling System Actuation Instrumentation can provide the initiating actions to mitigate the consequences of accidents that are beyond the ability of the operator to control.

Tables 3.3.3-1, 3.3.3-2, and 3.3.3-3 form the bases to ensure the effectiveness of the instrumentation used to initiate the actions. The proposed action does not affect the required number of operable channels (Table 3.3.3-1), the setpoints (Table 3.3.3-2),

or the response times (Table 3.3.3-3).

SER NO.: 92-132 CROSS

REFERENCE:

GO-100(200)-006; Units ¹1 & ¹2 DESCRIPTION OF CHANGE:

Change plant operating procedure to allow the reactor vessel to be completely flooded to 343" as opposed to the current level of 240"-265".

SUMMARY

No. If the reactor vessel is completely filled with water (level 343") and CRD flow is greater than 75 gpm, or either valves HV-141(241)-001, or HV-141(241)-002 is closed, the vessel could pressurize to 98 psig which would isolate RHR shutdown cooling. The pressurization could be controlled by RWCU, which is operable, operated in blowdown mode. If RWCU were to fail, two (2) SRV's would be operable (allowing for a single failure) to prevent any further pressurization. The SRV's will remain operable if water is vented through them (NEDE-24988-P). These two (2) means of pressure control ensures that the design pressure of 1400 psig (FSAR Section 5.3.2.1) will not be reached, and also reduces the possibility of pressurizing past 98 psig which would isolate RHR shutdown cooling.

No. A review of FSAR Chapters 4 and 5 indicates that the possible malfunctions that could occur due to the proposed action have been previously addressed. The method of currently purging the reactor vessel of non-condensible fission products is by the Standby Gas System, the proposed action would remove the fission products in the same manner, thus there is no different accident scenario or malfunctions that will cause a radioactive release.

No. Technical Specification 3/4.4.6 requires the reactor be maintained within certain pressure and temperature boundaries, with two (2) operable SRV's and RWCU operable, the reactor vessel temperature and pressure will remain to the right of Curve A which is shown in Figure 3.4.6.1-1 of the Technical Specifications.

SER NO.: 92-133 CROSS

REFERENCE:

DCP 92-9030; Unit ¹2 DESCRIPTION OF CHANGE:

Replace the present HYCAL RTD's during a Unit ¹2 RIO with new Conax Buffalo RTD assemblies that are not prone to moisture induced failure and will not utilize asbestos containing material.

SUMMARY

No. With the increased reliability of the replacement RTD, the probability of occurrence of SPOTMOS inoperability should decrease. FSAR Section 7.6.1b.12 and Chapters 6 and 15 were reviewed for this evaluation.

No. Potential failure modes of the replacement RTDs (open, short and drift) are the same as the existing RTDs and would be detected by the electronics unit or by the plant computer system allowing the plant operator to remove the failed sensor from the temperature average calculation.

No. The Conax Buffalo RTD will upgrade the SPOTMOS system by increasing the availability of suppression pool temperature data. There is no change to the quantity or location of sensors, or the redundant features of the SPOTMOS system that could affect any safety margins. Technical Specifications 3.3.7.4, 3.3.7.5 and 3.6.2.1 were reviewed for this analysis,

SER NO.: 92-134 CROSS

REFERENCE:

DCP 92-9027; Unit 02 DESCRIPTION OF CHANCE:

Installation of a milliamp meter in series with the existing Main Generator Field ground relay on Panel 2C-103D in place of existing bypass to monitor generator ground current.

SUMMARY

No. The design of the ammeter is such that it is durable in construction and due to the electromagnetic inductive nature of sensing electrical current where moving parts is held to a minimum, the probability'of ammeter malfunction is negligible. FSAR Chapters 6 and 15 were reviewed.

No. The proposed action does not affect the operation of the main generator. The proposed change will provide a permanent method for the monitoring of generator ground current.

Failure of the ammeter will not prevent the generator from performing it's function.

Accident scenarios have previously been analyzed in FSAR Chapters 6 and 15.

No. The proposed modification does not interfere with the logic, control or operation of any safety related plant system or component. There are no specific Technical Specification sections that apply to this modification.

SER NO.: 92-135 CROSS

REFERENCE:

TP-069-040; Unit - Common DESCRIPTION OF CHANCE:

LRW Collection Tank Organic Reduction Using Hydrogen Peroxide

SUMMARY

No. The proposed action involves injection of test chemicals to the LRW Collection Surge System and subsequent processing through the LRW filters and demineralizer to the sample tanks. The proposed test does not increase the radioactive inventory in any radwaste tank and does not involve the evaporator processing stream including the evaporator concentrates waste tank and associated piping and support equipment. FSAR Sections 15.7.3 and 2.4.13.3 were reviewed for this analysis.

No. Since the test does not affect the radioactivity contained in any LRW system, a failure as a result of the test would be no different than any possible failure. FSAR Section 15.7.3, involving failure of the evaporator concentrates tank, bounds all LRW integrity failures.

No. The proposed LRW processing change is intended to remove organic material from the liquid radwaste stream and does not involve (or change) the radioactivity that is stored or transferred in any waste stream. Technical Specification Sections 3/4.3.7, 3/4.4.4, 3/4.11, 3/4.12 were reviewed.

SER NO.: 92-136 CROSS

REFERENCE:

Bypass; Units ¹1 5 ¹2 DESCRIPTION OF CHANGE:

Installation of a temporary HEPA filter unit to the Containment Purge supply duct during plant outages to maintain a negative pressure inside primary containment.

SUMMARY

No. The Primary Containment Isolation function discussed in Section 6.2.4 is being affected by Bypass in that the NO/FC Primary Containment Isolation Valve HV-15721 (25721) will be placed in its failed closed position. Also, NO/FC Valves HV-15722, 23, 24, 25 (HV-25722, 23, 24, 25) will be maintained in their open position. This is acceptable since primary containment is not required in Conditions 1, 2, or 3. FSAR Section 9.4.2.1.1(f),

9.4.2.1.2 and 6.2.3 were also reviewed for this analysis.

Plo. Installation of this Bypass will preclude the possibility for an accident or malfunction of a different type than any previously evaluated in FSAR Sections 6.2.3, 6.2.4, 6.5.1, 9.4.2, and 15.0.

No. The only Technical Specification Section affected by this Bypass is Section 3.6.3, Primary Containment Isolation since Valve HV-15721 (25721) is affected. This is acceptable since the valve is being deenergized in the closed position.

SER NO.: 92-137 CROSS

REFERENCE:

DCP 92-9043; Unit ¹1 DESCRIPTION OF CHANGE:

Increase the elevation difference between the switch set and reset points of the drain line and the drain line of the Reactor Core Isolation Cooling System. This solution will allow the level switch to control the bypass valve as a backup to the steam trap in accordance with the original intent and to provide supervision of the condition of the steam trap via its alarm function.

SUMMARY

No. The performance of the RCIC steam line drainage will be improved by eliminating the current condition whereby steam is blowing down to the condenser most of the time while retaining the original maximum allowable drain pot level of 2 inches below the bottom of the steam line. This may contribute to a reduction in the concentration of feedwater metals which would have a beneficial effect on long term fuel performance. Sections 5.4.6, 6.2, 6.3, 15.4.8 and 15.6 of the FSAR were reviewed for this analysis.

No. The accidents of Sections 6.2, 6.3, and Chapter 15 were reviewed. No new operating modes are created for any system. The use of appropriate design, material, and construction standards ensures the integrity of the pressure boundary is equivalent to the original. The use of original criteria in the selection of level switch set and reset points assures functionality equivalent to the original design intent.

No. The Technical Specifications that are applicable include the following references to RCIC:

a brief mention of RCIC operability status relative to the ECCS systems in 3.5.1, Action c.,

identification of containment isolation valves in Table 3.6.3-1, and operability and surveillance requirements for the system in Section 3.7.3. The bases for these sections were reviewed for mention of any bases related to the drain pot of the steam supply in general. FSAR and SER Sections 5.4.6, 6.2, 6.3, and Chapter 15 were reviewed for design basis requirements related to the drain trap or steam supply that involve margin of safety.

No margins of safety are affected.

SER NO.: 92-138 CROSS

REFERENCE:

EO-100/200-103; Units ¹1 5 ¹2 DESCRIPTION OF CHANGE:

The purpose of this action is to eliminate the requirement to depressurize the RPV based upon the Heat Capacity Temperature Limit (HCTL), the associated Heat Capacity Level Limit (HCLL), or the HCTL segment of the Pressure Suppression Pressure (PSP) during ATWS events. This is done because calculations indicate severe damage may occur as a result of RPV depressurization during ATWS.

SUMMARY

No. This change only applies to ATWS events with failure of either SLCS or RHR which are beyond the design basis.

No. In fact, eliminating this requirement is designed to eliminate an accident of a different type than analyzed in the FSAR.

No. This action only applies to ATWS events which are beyond the design basis.

SER NO.: 92-139 CROSS

REFERENCE:

EO-100/200-103; Units ¹1 & ¹2 DESCRIPTION OF CHANGE:

RPV depressurization, when the containment pressure cannot be maintained below the Pressure Suppression Limit (PSL), is being restricted when the reactor power exceeds 5% and when the containment pressure exceeds the PSL. The purpose of the PSL is to ensure the pressure suppression function of the primary containment is maintained while the RPV is at pressure.

SUMMARY

No. This proposal only applies to ATWS events with the power in excess of 5%. These events are all beyond the design basis.

No. See I. above.

No. The PSL is not covered by Technical Specifications.

SER NO.: 92-140 CROSS

REFERENCE:

EO-1001200-103, Units ¹1 8 ¹2 DESCRIPTION OF CHANGE:

The Suppression Pool Design Temperature (SPDT) will be removed from the definition of the Heat Capacity Temperature Limit (HCTL). The original concern for imposing the SPDT has been resolved so that this limit is no longer necessary.

SUMMARY

No. The HCTL is a limit encountered only after the accident or malfunction has occurred. The accidents presented in Appendix I of the DAR bound the SRV while allowing the suppression pool temperatures to increase. These accidents were examined and in no case was the HCTL encountered. The quencher loads actually decrease as the pool temperature rises. This action actually reduces the consequences of an accident.

No. On the basis of analysis and experiment, it appears necessary to assign a limit for the suppression pool temperature based on quencher operation when the PP&L T-quencher is used for steam discharge. Thus, this action actually improves the containment performance.

No. The SPDT is covered in the Technical Specification Section 5.2.2. There are no actions specified. It is referenced as a design feature only. There is no action feature associated with it.

SER NO.: 92-141 CROSS

REFERENCE:

EO-100/200-103; Units ¹1 & ¹2 DESCRIPTION OF CHANGE:

The Primary Containment Control Procedure is being upgraded to the BWROG EPG Revision 4.

SUMMARY

No. The actions in this procedure are taken after the incidence of the initiating event. This procedure implements all of the operator actions in the FSAR or the DAR which are assumed to occur when demonstrating compliance of the containment design with the regulations. The actions identified are called out in the procedures and are within the design basis. Additionally, actions are called out to restore the containment parameters to the values assumed in the safety analysis.

No. The actions directed by this procedure for conditions that are in the design basis are designed to either implement the actions specified in the safety analyses described in Chapters 6 and 15 of the FSAR or restore the containment process parameters to normal values. Actions which are inconsistent with the design bases analysis are only implemented when the accident has progressed beyond the design bases.

No. The operator actions proposed do not involve changing any Technical Specification basis.

SER NO.: 92-142 CROSS

REFERENCE:

EO-100/200-113; Units ¹1 8 ¹2 DESCRIPTION OF CHANGE:

Upgrade of the Susquehanna Emergency Operating Procedures to Revision 4 of the BWROG Emergency Procedure Cuidelines.

SUMMARY

No. Although Section 15.8 of the FSAR presents a brief discussion of ATWS, this accident is beyond the plant design basis and therefore this section of the Safety Evaluation is not applicable.

No. The proposed actions are only executed after an ATWS event has occurred. Therefore, these actions are only executed when the plant is in a configuration which is beyond the design basis. Consequently, this section of the Safety Evaluation is not applicable.

No. The proposed actions do not affect any Technical Specification requirements. Operating procedures only specify operator actions that would be carried out with the plant in a configuration which is beyond the design basis.

SER NO.: 92-143 CROSS

REFERENCE:

DCP 91-3021, ME-2RF-003; Unit ¹2 DESCRIPTION OF CHANGE:

Installation of clamps on six jet pump instrument sensing lines. The purpose is to prevent the instrument sensing lines from failing and will allow removal of this failure mechanism's contribution to the 88%

recirculation pump speed administrative limit.

SUMMARY

No. All equipment is being used within its design basis except that six jet pumps are being disassembled and a rod block is being jumpered out.

FSAR Section 5.4.'I.4 specifies that the jet pump assembly provides a floodable volume of 2/3 core height; however, this work is being performed with the core offloaded and therefore the core height is zero.

The FSAR Chapter 6 and 15 events, especially the Decrease of Reactor Coolant Inventory sections, are evaluating situations when fuel is in the reactor. Since no fuel will be in the vessel when this activity will be performed, there are no negative effects on safety.

No. The core will be defueled during the activity, no new failure modes are created, and the plant will be in an improved condition after the modification is installed.

No. The jet pump operability Technical Specification 3/4.4.1.2 is only required during Conditions 1 5 2. Since all this work is being completed with the core defueled, no decrease in any margin of safety will occur.

SER NO.: 92-144 CROSS

REFERENCE:

TP-237-006; Unit 02 DESCRIPTION OF CHANCE:

Providing an alternate source of keepfill for Unit 02 Core Spray from demineralizer water to maintain the system operable during the Unit ff2 refuel outage when the vessel is defueled in Condition 5.

SUMMARY

No. Per FSAR Section 6.3.2.2.5, ECCS discharge line fill system ensures the system is full by maintaining a pressure greater than atmospheric at the system high points to prevent air accumulation. The FSAR also estimates leakage at less than one gpm. The demineralizer water system has a redundant pump in case the lead pump fails. This makes the demineralizer system as reliable a source of keepfill as the Condensate Transfer system which is also non-seismic, non-quality, and non-ASME.

This bypass does present new failures such as the hose and human error potential for closing the demineralizer valves and the Core Spray high point valves. However, this minor increase in risk is offset by the controls which will be placed on the bypass.

To ensure there is no cross-contamination into the demineralizer water system, there will be 2 check valves installed on the hoses to prevent backflow from Core Spray into the hoses. To minimize the time the keepfill pressure is low, should that occur, TP-237-006 will control the installation of the bypass and monitor keepfill pressure while the bypass is installed.

No. Condensate Transfer normally is the source of water for the keepfill system. By providing keepfill from demineralizer water rather than Condensate Transfer for less than 2 days, the risk factor is only slightly increased. The hose could fail, the hose drop valves or the high point valves could be closed or the demineralizer system could fail. But, as discussed earlier, the loss of keepfill by a failure of the demineralizer system would be no different than a loss of keepfill due to a Condensate Transfer system failure. This is in accordance with FSAR Section 6.3.2.2.5.

No. In Condition 5 per Technical Specification 3.5.2, Core Spray subsystems of 2 operable pumps and an operable flowpath from the suppression chamber or from the CST to the Reactor vessel must be operable. However, the note for Condition 5 states that the ECCS is not required to be operable if the vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed and water level is sufficient per Technical Specifications 3.9.8 and 3.9.9. Since these conditions will be met, the Core Spray subsystems are not required to be operable either, per Technical Specifications.

SER NO.: 92-145 CROSS

REFERENCE:

DCP 92-9071A, B, C, D; Unit Common DESCRIPTION OF CHANGE:

Add branch drain lines from each end of the header pipe that will tie into the existing drain tubing downstream of the drain header in the Emergency Diesel Generator fuel oil drain system. Additional venting will be added for each header pipe by providing a vent at the beginning of the middle drain leg. This modification will further improve the flow of fuel from the fuel injector pumps to the day tank overflow line of the Emergency Diesel Generators.

SUMMARY

No. The SAR has been reviewed; specifically Chapter 15, "Accident Analysis" and Chapter 9, Section 9.5, "Other Auxiliary Systems."

This modification consists of adding additional venting and drains to the header pipe and reconfigure the drain to the overflow line to eliminate as much as possible any standing column of oil. All these changes are to improve the gravity draining capability of the system.

This modification will not change the function of the system but will actually enhance its ability to operate as intended.

No. The new vent and drain lines added will conform to the same requirements, both for

~

installation and service, as the existing fuel oil tubing. All tubing and supports are classified as Seismic Category I and will be designed to meet these seismic requirements.

The actual operation of the fuel oil drain system will be improved by providing better flow through the injection pump pedestal to the fuel oil storage tank.

No. The Technical Specifications have been reviewed, specifically Section 3/4.8, "Electrical Power Systems." Since the original piping codes and standards are utilized, the implementation of this modification will not change the function or adversely affect the operation of any component or system related to the emergency diesel generators. The addition of these vents and drains to provide better flow through the drain system will only improve the system reliability and ability to operate as intended.

SER NO.: 92-146 CROSS

REFERENCE:

DCP 92-9006; Unit ¹2 DESCRIPTION OF CHANCE:

Change the cleanup line isolation signal from reactor vessel low water level 3 (+13") or high drywell pressure, to level 2 (-38") or high drywell pressure. This increases the margin between water levels during power operation and the isolation level and therefore increases the likelihood that the cleanup line will be available following transients.

SUMMARY

No. The isolation signal change from level 3 to 2 does not degrade the operation of any equipment. The relays and level switches to be used for the level 2 isolation signals are identical in design, material and construction to those currently used for level 3. FSAR Sections 6.2, 6.3 and Chapter 15 were reviewed.

No. The modification uses spare terminals on existing relays to receive an input from a level 2 switch. These relays are identical to those used to receive the level 3 input signal. The actuation logic remains single failure proof. The change in internal wiring is in accordance with existing installation and termination procedures. The use of the cleanup line serves to lower the pool water level which decreases the loading during a LOCA or safety/relief valve operation. This is in accordance with FSAR Sections 6.2, 6.3 and Chapter 15.

No. The reactor vessel low water level signal for isolation of the suppression pool cleanup is contained in Technical Specification Table 3.6.3-1. It indicates level 3 for Unit ¹2. This is not explicitly stated in the bases of the Technical Specification.

SER NO.: 92-147 CROSS

REFERENCE:

DCP 92-9015; Unit ¹2 DESCRIPTION OF CHANCE:

New decontamination equipment has been installed in the Decontamination Shop in Area 42, Elevation 676'f the Radwaste Building. A 52 gallon hot water heater will be added. Installation of new 120V AC receptacles will eliminate extension cords and tripping hazards. The Electropolisher Assembly power feeders will be disconnected and removed.

SUMMARY

I. No. The deletion/addition of new equipment in the Radwaste Decontamination Shop does not adversely affect safety related systems or equipment. Each new piece of equipment is protected by a circuit breaker in panel OPP-313 and internal fuses in each control cabinet.

This does not interfere with the logic, control or operation of any safety related plant systems or components. FSAR Chapters 6 and 15 were reviewed.

No. The proposed change will provide a permanent method for decontaminating tools and equipment. Also, this modification will add a hot water heater for the decontamination shower and sinks. Failure of this equipment will not prevent the plant's safety equipment from performing their function. Accident scenarios have previously been evaluated in FSAR Chapters 6 and 15.

n, No. There are no specific Technical Specification sections that apply to this modification.

SER NO.: 92-148 CROSS

REFERENCE:

DCP 92-9072; Unit t2 DESCRIPTION OF CHANCE:

This modification will resolve NCR No.92-020 by replacing a SOR pressure switch containing a Kapton diaphragm with a nuclear qualified SOR pressure switch containing a stainless steel diaphragm. Also, the orientation of the switch pressure port will be changed to minimize the affects of vibration.

SUMMARY

No. Review of the SAR indicates the only accident potentially affected by this switch failure is Increase in Reactor Pressure caused by loss of condenser vacuum, FSAR Section 15.2.5.

This event is categorized as an accident of moderate frequency. There are no common mode failures introduced. The new switch and terminal box will be installed in accordance with requirements as discussed in the Safety Impact Item Evaluation. RIE 92-0016 has demonstrated that the Nuclear Qualified SOR Inc. switch and the Barksdale switch are equivalent.

No. Replacement of the 'B'witch with a Nuclear Qualified SOR Inc. vacuum switch does not alter the function or operation of the switch. The evaluation does not indicate any new failure mode for the new switch.

No. Technical Specification Bases Section 3/4.3.2, Isolation Actuation Instrumentation, ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of reactor systems.

Except for the MSIVs, the safety analysis does not address individual sensor response times.

Per Technical Specification 3.3.2-3.3.e, the response time for Condenser Vacuum Low is not applicable.

SER NO.: 92-149 CROSS

REFERENCE:

DCP 90-3036, Unit 02 DESCRIPTION OF CHANGE:

The intent is to replace the existing RWCU pumps with new sealless pumps, each with at least 100% system capacity.

SUMMARY

No. FSAR Sections 3.6.2.1.1.b)7), 5.4.8.1, 2 & 3, 7.3.1.1A.2.4.1.9 & 10, 9.2.10 and 3.8.4.2, and Tables 3.2-1 and 5.4-2 were reviewed for applicability. The proposed modification does not alter the function of the Reactor Water Cleanup System.

No. The replacement pump, piping, purge water system, and other associated components are in compliance with FSAR criteria and regulatory requirements and the leakage detection capability is not affected.

No. Technical Specification Sections 3.3.2, 4.3.2.1, 4.3.2.2 and 4.3.2.3 are applicable to Reactor Water Cleanup System. Table 3.3.2-2 lists the isolation actuation instrumentation trip setpoints for the system. These sections and their implied basis were examined and it was determined that the margin of safety as defined in this basis is not affected by this modification. Furthermore, isolation actuation instrumentation and setpoints applicable to Reactor Water Cleanup System are unchanged. No accident analysis is affected by this modification.

SER NO.: 92-150 CROSS

REFERENCE:

DCP 92-3021; Unit 42 DESCRIPTION OF CHANGE:

Replacement of the worm shaft gear and motor pinion gear in the actuator on HV-251F004A. This will satisfy the requirements of both the reduced voltage pullout and the LLRT. This is performed in the RHR system.

SUMMARY

No. After implementation of this modification, the MOV time will be increased from 123 seconds to 200 seconds. This has no adverse affect on the valve's safety function because it does not receive any automatic isolation signals. FSAR Sections 15.6 and 3.9.3.2b.2 were reviewed for this analysis.

No. This modification does not affect the ability of RHR to perform its design intended function as described in FSAR Sections 5.4.7 and 6.3.2.2.4. It does not affect any system requirements for a specific stroke time. It does not alter the MOV's allowable seismic acceleration. It also does not affect any other equipment.

No. The bases for Technical Specifications 3/4.3.2 and 3/4.3.3 discuss reactor system isolation actuation instrumentation and ECCS system actuation instrumentation. Since the modification affects no equipment other than HV-251F004A, these bases are unaffected.

The modification has no effect on LPCI operation since the valve is normally open and therefore does not necessitate a change to the bases for Technical Specification 3/4.5, "Emergency Core Cooling Systems." Alignment of the RHR system from LPCI to shutdown cooling and back is performed as a manual operation. The modification has no effect on shutdown cooling operation or realignment to LPCI since the valve cycle is manually initiated, and therefore does not necessitate a change to the bases for Technical Specification 3/4.4.9, "Residual Heat Removal." With regard to the containment isolation function of HV-251F004A, the integrity of the valve is not adversely affected and it remains under remote manual control without a specified closure time per Table 3.6.3-1; therefore, the bases for Technical Specification 3/4.6.1, "Primary Containment," and Technical Specification 3/4.6.3, "Primary Containment Isolation Valves" are unaffected.

SER NO.: 92-151 CROSS

REFERENCE:

DCP 92-9073; Unit ¹2 DESCRIPTION OF CHANGE:

The components affected by this modification are feedwater check valves HV-241F032A & B. The eight cap screws are being replaced because they were broken. They will be tack welded into position and be made of "Q1" austenitic stainless steel. This will increase the tensile stress area of the screws by 3.7 times. The tack weld will prevent the screws from backing out.

SUMMARY

No. FSAR Sections 6.2.4, 6.2.4.3.2.1, 6.2.3.2.1, 5.4.9, Table 6.2.12, Chapter 15 were reviewed for this evaluation. The design and construction standards are not changed.

The new bolting material is in common use in the plant. The basic functions of any plant system are not changed. The change to the seat retaining ring and disc will have no affect on valve performance. This reduces the likelihood of seat retaining ring screw failure.

No. This modification creates the possibility for interferences that prevent the valves from opening or closing due to the increased disc outside diameter. However, adequate clearance for both installation and operation were confirmed by the vendors review of the shop fabrication drawings and will be reconfirmed by exercising the modified valves and observing internal clearances as part of the requirements of the change.

Therefore, the risk of this type is considered negligible.

No. Technical Specification Basis 3/4.6.1.1 ensures that the release of radioactive materials from containment is restricted to the leak paths and rates assumed in the accident analyses. Technical Specification Basis 3/4.6.1.2 ensures that the total containment leakage value will not exceed the value assumed in the accident analyses. Lastly, Technical Specification Basis 3/4.6.3 ensures that the containment atmosphere will be isolated from the outside environment in the event of radioactive release or containment pressurization.

SER NO.: 92-152 CROSS

REFERENCE:

DCP 92-9010; Unit ¹1 DESCRIPTION OF CHANGE:

PP&L NCR 92-019 was generated to document overfilled/overweight cable trays. The modification is to add an additional cable tray support.

SUMMARY

No. This modification takes place in Unit ¹1 Reactor Building on 749'able tray section E1KK16/17 and does not interfere with the logic, control or operation of any safety related plant systems or electrical equipment. FSAR Chapters 6 and 15 were reviewed.

II I.'o. No.

Thermo-Lag will be reinstalled on tray and support per Technical Specification 3/4.7.7.

When cable support tray is added, Thermo-Lag will be removed and replaced per Technical Specification 3/4.7.7.

SER NO.: 92-153 CROSS

REFERENCE:

DCP 92-9025; Unit ¹1 DESCRIPTION OF CHANCE:

The installation of two air louvers in each Branch Junction Module (BJM) enclosure to provide natural draft ventilation for all component parts in the enclosure.

SUMMARY

No. The improved cooling accomplished by this modification will provide for increased dependability regarding Reactor Manual Control System (RMCS) operation and therefore increased Unit ¹1 reliability. Failure of the louvers to perform properly would not prevent the RMCS equipment from performing its design function. A review of the accidents described in FSAR Chapters 6 and 15 was done.

No. This modification will be engineered in accordance with existing standards and procedures. No safety related circuits will be affected. The proposed action increases the reliability of the RMCS by providing BJM ventilation consistent with its intended design and function. The proposed modification does not involve changes in system operation nor add a more severe or different type of failure mode. Loss of the RMCS does not impair the reactor's ability to scram which ensures a safe condition for the plant.

No. Technical Specification Section 3/4.1 describes the operational requirements of the Reactivity Control Systems. This modification will only enhance the performance of certain equipment related to reactor control. The ability to insert/withdraw control rods as originally designed will not be diminished. The proposed modification will not impair the operation of any equipment or power supplies needed for safe shutdown or control of accident conditions. Technical Specification 3/4.1.4.1 and 3/4.1.4.2 delineates specific reactivity control operability and surveillance requirements. The proposed modification does not adversely impact RMCS operability or surveillance criteria.

SER NO.: 92-154 CROSS

REFERENCE:

DCP 92-9074; Unit 02 DESCRIPTION OF CHANGE:

The signal cable for the Unit II Local Power Range Monitor (LPRM) ff24-33D detector assembly has been damaged under the vessel. This modification provides for a plug jack connection in the undervessel cable tray and for new cabling from the plug jack to the SMA jack connector for LPRM ff24-33D.

SUMMARY

No. This modification provides for a plug jack connection in the undervessel cable tray and for new cabling from the plug jack to Unit C2 LPRM ff24-33D. The plug jack connector will be placed and ty-wrapped in a seismically supported cable tray with heat shrink tubing being applied to totally enclose the connector and, as such, no potential safety impact concerns can result. No degradation in signal quality will occur. The cabling is electrically equivalent and the new cable connectors are completely compatible with the existing installation.

No. ~ The LPRMs are not environmentally qualified pieces of equipment because the LPRMs perform their design function in triggering required RPS actuations and are not required after any accident conditions occur. Therefore, the plug jack connection proposed by this modification need not be documented as an EQ qualified connection.

No. Technical Specification Basis 3/4.3.1, "Reactor Protection System Instrumentation,"

specifies the minimum operability and surveillance requirements involving nuclear instrumentation. Similarly, Technical Specification 3/4.3.6, "Control Rod Block Instrumentation" and Technical Specification 3/4.4.1, "Reactivity Control Systems" has been reviewed. The proposed modification has no impact on the above mentioned Technical Specifications. The proposed modification does not adversely affect the logic, control, or operation of any safety related plant system or component. The performance characteristics between the modified LPRM cabling and the existing LPRM cabling are essentially identical.

SER NO.: 92-155 CROSS

REFERENCE:

DCP 92-9075; Unit ¹2 DESCRIPTION OF CHANGE:

This modification removes the space heaters from service for motor operated valves (MOVs). This is applicable to MOVs of the following systems: Reactor Building Closed Cooling Water, Containment Instrument Gas, Feedwater, Residual Heat Removal, Reactor Core Isolation Cooling, Control Rod Hydraulics, High Pressure Coolant Injection, Reactor Water Clean-up, Reactor Vessel and Auxiliaries, Reactor Recirculation and Main Steam. The reason for this is the outer jacket of each conductor had deterioration on the entire length of cable inside MOV housing.

SUMMARY

No. The electrical operability of the MOVs are not affected if the space heaters are deenergized. This modification actually results in a decrease in the probability of occurrence of malfunctioning of equipment. The removal of the space heaters from service has no affect on any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR.

No. The deenergizing of the space heaters would not create the possibility for an accident of a different type which is in accordance with FSAR Chapters 6 and 15. The modification is in accordance with existing approved installation procedures.

Electrical separation is maintained in accordance with FSAR Section 8.3.1.11.4.

No. There is no affect on the operability of the MOVs. The proposed action associated with these valves does not affect the operability requirements, surveillance requirements, or any existing margin of safety in any Technical Specification.

SER NO.: 92-156 CROSS

REFERENCE:

DCP 92-9006; Unit ¹2 DESCRIPTION OF CHANCE:

Change the cleanup line isolation signal from reactor vessel low water level 3 (+13") or high drywell pressure, to level 2 (-38") or high drywell pressure. This increases the margin between water levels during power operation and the isolation level and therefore increases the likelihood that cleanup line will be available following transients.

SUMMARY

No. The isolation signal change from level 3 to 2 does not degrade the operation of any equipment. The Agastat relays and level switches to be used for the level 2 isolation signals are identical in design, material and construction to those currently used for level 3. FSAR Sections 6.2, 6.3 and Chapter 15 were reviewed.

No. The modification uses spare terminals on existing relays to receive an input from a level 2 switch. These relays are identical to those used to receive the level 3 input signal. The actuation logic remains single failure proof. The change in internal wiring is in accordance with existing installation and termination procedures. The use of the cleanup line serves to lower the pool water level which decreases the loading during a LOCA or safety/relief valve operation. This is in accordance with FSAR Sections 6.2, 6.3 and Chapter 15.

No. The reactor vessel low water level signal for isolation of the suppression pool cleanup is contained in Technical Specification Table 3.6.3-1. It indicates level 3 for Unit ¹2. This is not explicitly stated in the bases of Technical Specification.

SER NO.: 92-157 CROSS

REFERENCE:

DCP 92-9078; Unit t2 DESCRIPTION OF CHANCE:

This modification corrects a nonconforming condition and allows pipe whip restraints PR-211A and PR-220A to perform their original design function; to protect other safety related components from being damaged during a main steam line pipe rupture.

SUMMARY

No. FSAR Section 3.6, "Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping," and in particular, Subsection 3.6.1.2.1, "Main Steam System,"

describe pipe whip restraints such as PR-211A and PR-220A as having only one function. That function is to prevent damage to safety related components essential to reactor safe shutdown in the event of a main steam pipe rupture. This modification ensures proper gaps on the pipe whip restraints PR-211A and PR-220A and will validate the original design basis. FSAR Chapters 6 and 15 have been reviewed.

No. This modification will enable pipe whip restraints PR-211A and PR-220A to perform their original design function of restraining the main steam piping during a postulated pipe rupture and preventing any damage to safety related components or equipment.

This modification does not deviate from the original design basis.

No. This modification corrects a nonconformance.

SER NO.: 92-158 CROSS

REFERENCE:

DCP 92-9077; Unit 02 DESCRIPTION OF CHANCE:

To ensure no contact between the main steam relief valve bonnet and the restraint, a section of steel plates making up the restraint, is coped out approximately 1" in the direction of the valve's thermal movements at the area where contact was made.

SUMMARY

No. The SAR has been evaluated, specifically Sections 3.6, "Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping," and 15.0, "Accident Analysis."

Section 3.6 of the FSAR addresses the consequences of a high energy line break and describes the required protection to guard against such pipe failures. The whip restraint is the means of protection in preventing pipe whip damage to safety related equipment/components essential to reactor safe shutdown. The new configuration of pipe restraint PR-19 will not be degraded and will be capable of performing its intended function. In providing this clearance, the main steam line will conform to its as-analyzed condition, eliminating any additional thermal stresses caused by the interference.

No. Modifying pipe restraint PR-19 to provide the required thermal clearances enhances the present situation and eliminates the nonconforming condition. The modification will bring the main steam system back to its as-analyzed condition while maintaining the pipe restraint's design function.

. I II. No. The Unit 42 Technical Specifications have been reviewed and pipe whip restraints are not specifically addressed. The qualification of the pipe whip restraint ensures that providing additional clearance will not degrade the restraint from performing its design function of protecting safety related components from damage due to a main steam pipe rupture.

SER NO.: 92-159 CROSS

REFERENCE:

EO-100/200-112 5 113; Units ff1 8 ff2 DESCRIPTION OF CHANCE:

A deviation will be taken in the implementation of BWR Owner's Group (BWROG) Emergency Procedure Guideline (EPG) regarding the use of HPCI during a Reactor Pressure Vessel (RPV) depressurization, under ATWS conditions.

SUMMARY

No. The proposed deviation allows the use of a plant safety system (HPCI) to provide core cooling during an ATWS depressurization transient. This is consistent with the SSES FSAR Section 15A.6.6.3 which identifies the concurrent use of both HPCI and RCIC as an appropriate RPV level control approach during ATWS. The result of this deviation is that core cooling is enhanced, and the risk of core damage is reduced.

I The proposed deviation will result in a slightly higher ATWS power level than would be the case if the BWROG EPG were followed. The magnitude of this slight power level increase has been calculated to produce a change in suppression pool temperature of roughly 6'F. This small increase in suppression pool temperature is judged to be inconsequential because the temperature of the suppression pool at the onset of the event can vary by as much as 10'F or more. In addition, ATWS power levels can be such that suppression pool temperature will rise by 6'F in roughly 2 minutes time.

No. The proposed deviation only comes into play when an ATWS has already occurred.

In that event, the HPCI system's use is permitted, though not required, to ensure adequate core cooling. Failure of the HPCI system during its use as proposed results in the same system configuration as currently required by the BWROG EPG. As a result, additional operational flexibility is provided to further ensure adequate core cooling during the ATWS event. As has been stated previously, the proposed use of HPCI enhances core cooling while producing an insignificant additional containment heat loading. The effects of this additional containment heat load is expected to reduce the time to reaching containment limits by roughly 2 minutes.

No. The proposed change has no impact on Plant Technical Specifications.

The proposed change simply allows the use of HPCI during the depressurization portion of the ATWS transient. Technical Specifications (3.5.1.C) currently provide requirements regarding HPCI operability. The proposed change does not affect these requirements.

SER NO.: 92-160 CROSS

REFERENCE:

EO-100/200-105; Units C1 5 f)2 DESCRIPTION OF CHANCE:

This change is to support the proposed revision of EO-100/200-105 (Radioactivity Release) to incorporate Rev.

4 of the BWROG Emergency Procedure Guidelines (EPG).

SUMMARY

No. The action to manually initiate ADS when a primary system is discharging to areas outside of primary and secondary containments and release rates approach those associated with the declaration of a General Emergency is consistent with the use of ADS as described in Section 7.3.1.1A.1.4 of the FSAR which acknowledges the manual operation of the ADS system. The plant transient which results from an ADS actuation has been analyzed in the FSAR and is less severe than other analyzed transients such as the Large Break LOCA Transient (FSAR Table 3.9-15).

No. The actions specified are only taken in response to plant conditions involving the break and failure to isolate of a primary system which results in large releases of radioactivity to areas outside of Primary and Secondary Containment. Conditions such as these are only anticipated after the failures of multiple plant components including the affected primary system pressure boundary and the affected primary system isolation valves. The actions specified are mitigative in nature and will reduce the consequences of this unlikely sequence of events.

No. The actions specified requires the use of designated plant equipment (TB MVAC/RPS/ADS) in a manner which is consistent with the intended use of that equipment and with FSAR Analysis Assumptions regarding the use of the equipment.

In addition, as was mentioned previously, the actions associated with performing a reactor shutdown when a primary system is discharging to areas outside of Primary and Secondary Containments, (and cannot be isolated) is, itself, consistent with the action statement associated with the Technical Specification LCO for Reactor Coolant System Operational Leakage (T.S. 3.4.3.2) a well as the action statement for the Technical Specification associated with Primary Containment Isolation Valves (T.S.

3.6.3).

SER NO.: 92-161 CROSS

REFERENCE:

DCP 92-9079; Unit 02 DESCRIPTION OF CHANGE:

A .25" (approximately) seal weld will be applied at the seam around the entire flanged connection between the valve body and bonnet of HV-241F022B. The modification adds a nonstructural seal weld to the body-bonnet connection of the valve in accordance with the manufacturer's design.

SUMMARY

No. FSAR Sections 6.2.4, 5.4.9, Table 6.2.12 and Chapter 15 were reviewed for this analysis. These bases are not affected by this modification.

No. The modification adds a nonstructural seal weld to the bod y-bonnet connection o f the valve in accordance with the manufacturer's design. The option to use this weld was developed as part of the original valve design although it was not explicitly stated on the design drawings submitted for our purchase. Therefore, there will be no change in the performance of the valve after the weld is applied. There are no other components or systems affected by the modification.

No. Technical Specification Basis 3/4.6.1.1 ensures that the release of radioactive materials from containment is restricted to the leak paths and rates assumed in the accident analyses. Technical Specification Basis 3/4.6.1.2 ensures that the total containment leakage value will not exceed the value assumed in the accident analysis at peak accident pressure. Technical Specification Basis 3/4.6.3 ensures that the containment atmosphere will be isolated from the outside environment in the event of radioactive release or containment pressurization. Lastly, Technical Specification Basis 3/4.4.7 ensures containment integrity in case of a line break and the maximum closure time of the MSIV ensures that the core is not uncovered following a line break.

SER NO.: 92-162 CROSS

REFERENCE:

DCP 92-5006; Unit - Common DESCRIPTION OF CHANCE:

Installing a modification to the Security Data Management System (SDMS) software which will bypass the anti-passback restriction for predesignated personnel.

SUMMARY

No. The proposed action will enable essential personnel to quickly respond in time of emergency. Also, this will not alter the intent of the computer-based Security Data Management System or its performance capabilities set forth in Section 13.6 of the FSAR.

No. The proposed change will be implemented on the Security Computer System which does not interface with or affect any plant equipment. It does not affect the ability of the security computer to monitor potential security violations.

No. The proposed modification will be implemented on the Security Data Management System which is unable to exercise independent control of the plant. Plant operation will be unaffected by a modification to the SDMS software.

SER NO.: 92-163 CROSS

REFERENCE:

DCP 92-9082; Unit If2 DESCRIPTION OF CHANCE:

LPRM 32-33B signal cables will be relocated to an identical spare penetration connector because the current penetration connector is defective and cannot be repaired.

SUMMARY

No. This modification simply utilizes a neighboring spare connector in penetration 2W100A. Since the connectors are identical, no degradation in signal quality will occur. The connectors are electrically identical and compatible with the original installation.

No. The modified LPRM penetration connection is identical to the original configuration.

No. Technical Specification Basis 3/4.3.1, "Reactor Protection System Instrumentation,"

specifies the minimum operability and surveillance requirements involving nuclear instrumentation. Similarly, Technical Specification 3/4.3.6, "Control Rod Block Instrumentation" and Technical Specification 3/4 4.1, "Reactivity Control Systems" has been reviewed. The proposed modification has no impact on the above mentioned Technical Specifications. The proposed modification does not adversely affect the logic, control, or operation of any safety related plant system or component. The performance characteristics between the modified LPRM cabling and the existing LPRM cabling are identical.

SER NO.: 92-164 CROSS

REFERENCE:

DCP 92-9033; Unit ff2 DESCRIPTION OF CHANCE:

There are two pipe support frames with two supports on each frame that are rigidly attached to both the Containment Building and the Reactor Building platform steel. This change will modify the pipe support frames by removing the braces which attach to the containment wall and installing new bracing back to the Reactor Building platform steel.

SUMMARY

No. The SAR has been evaluated, specifically Sections 9.4.2, "Reactor Building Ventilation System," 15.0, "Accident Analysis," and 3.7b, "Seismic Design".

The pipe supports are required to support and restrain the ESW piping to the RHR Unit Cooler 2V-210B such that the cooler remains operational as outlined in Section 9.4.2 of the FSAR. Removing the attachment from the containment wall and redirecting the brace to the Reactor Building steel ensures compliance with the piping analysis and pipe support qualification. Additional differential building movement forces will be eliminated from the pipe support structure and from the structural platform steel. Note that FSAR Section 3.7b requires that each Seismic Category I structure be independent; by removing these braces we will be helping to ensure this condition.

No. Modifying the pipe supports as required by the disposition of EDR No. C00067 will eliminate the rigid attachments made between the Containment and Reactor Building and thus remove the additional loads imposed by a seismic event. The modification will bring the supports in conformance with its original design intent.

No. Section 3.7.1.2 of the Unit t2 Technical Specifications provides the Limiting Condition for Operation of the ESW System. The operability of the ESW System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. Removing the brace from the containment wall and installing it to the Reactor Building structure will help ensure this requirement be met.

SER NO.: 92-165 CROSS

REFERENCE:

DCP 92-9081; Unit ¹2 DESCRIPTION OF CHANGE:

Bypass Nos. 2-92-027 and 2-92-028 were initiated to temporarily rectify the introduction of an unwanted foreign potential into the TSH-24021A2, B2 circuitry and vice versa. It was also done to return the scheme logic to original design intent. The modification is to replace the temperature switch cable terminations at the terminal block with in-line cable splices, resulting in segregation of the 125 VDC and 24 VDC circuits.

SUMMARY

No. It is shown in Chapter 15 of the FSAR that none of the malfunctions result in significant fuel damage. The recirculation system has sufficient flow coastdown characteristics to maintain full thermal margins during abnormal operational transients.

All cable splices involving TSH-24021A2, B2 control circuits have already been installed in accordance with accepted plant procedures. The specific Reactor Recirculation MG set control circuits are wired properly and function according to their original design basis.

No. The proposed action does not affect any safety systems which insure the safe shutdown of the plant. Failure of the Reactor Recirculation MG sets will not prevent the plant safe shutdown equipment from performing their design safety functions.

Accident scenarios have previously been analyzed in FSAR Chapters 6 and 15.

No. Technical Specification Basis 3/4.4.1 "Recirculation System" states that two reactor coolant system recirculation loops shall be in operation with the reactor at specified thermal power/core flow conditions. The proposed modification does not adversely affect the logic, control or operation of any reactor coolant system recirculation loop.

The proposed modification enhances system performance by returning specific MG set control circuits into their design basis configuration.

SER NO.: 92-166 CROSS

REFERENCE:

EO-100/200-102; Units t1 5 42 DESCRIPTION OF CHANGE:

Changes to EO-100/200-102 that incorporate information from Rev. 4 of the Boiling Water Reactor Owner's Group (BWROG) Emergency Procedure Guideline (EPG).

SUMMARY

No. The direction to manually control ADS, the instruction to bypass the RCIC low steam pressure isolation logic and the instruction to bypass the HPCI suction auto swap logic are issues that were evaluated. A review of Sections 7.3.1.1A.1.4, 6.3.2.2.1, 6.2.1.1.3.3.1.1 and 6.2.1.1.3.1 of the FSAR was done. The consequences of any FSAR analyzed transient are not increased through the implementation of this proposed EOP action.

No. Implementation of EO-100/200-102 imposes tighter restrictions on the use of ADS than would be the case if transients associated with ADS actuations are less likely to occur inadvertently as a result of the implementation of EO-100/200-102.

Furthermore, an inadvertent ADS initiation produces a transient which is bounded by FSAR analyses. Failure to manually initiate ADS as required by Emergency Operating Procedures creates a condition in which subsequent EOP steps instruct the operator on many occasions to open SRV's. It is not expected that a trained crew of operators will continue to ignore procedural instructions and fail to manually initiate ADS, when directed. In addition, as previously stated, Section 7.3.1.1A.1.4 of the SSES FSAR acknowledges that manual control of ADS is an acceptable operating approach. The transient introduced resulting from the manual operation of ADS, is similar to, but is less severe than the FSAR-Analyzed Large Break LOCA Transient (FSAR Table 3.9-15).

No. The basis for Technical Specification 3.5.1 (ECCS) indicates that use of ADS (given HPCI failure) for small break LOCA will limit fuel cladding temperature to less than 2200'F. Use of ADS as specified in the BWROG EPG (which is the basis for EO-100/200-102, as well as other SSES EOP steps) also limits fuel cladding temperature to less than 2200'F.

SER NO.: 92-167 CROSS

REFERENCE:

EO-100/200-113; Unit ¹2 DESCRIPTION OF CHANCE:

Upgrade of the Susquehanna Emergency Operating Procedures to Rev. 4 of the BWROG Emergency Procedure Guidelines.

SUMMARY

No. Although Section 15.8 of the FSAR presents a brief discussion of ATWS, this accident is beyond the plant design basis and therefore this section of the Safety Evaluation is not applicable.

No. The proposed actions are only executed after an ATWS event has occurred.

Therefore, these actions are only executed when the plant is in a configuration which is beyond the design basis. Consequently, this section of the Safety Evaluation is not applicable.

No. The proposed actions do not affect any Technical Specification requirements. EO-100-113 only specifies operator actions that would be carried out with the plant in a configuration which is beyond the design'basis.

SER NO.: 92-168 CROSS

REFERENCE:

DCP 92-9084; Unit - Common DESCRIPTION OF CHANCE:

A spool piece will be installed in place of Check Valve 0-22-035, on the 14" Fire Protection Clarified Water Tank Supply piping, and blanks will be installed on the 4" inlet and outlet flanges to Domestic Water Storage Tank OT-517.

The addition of blanks to the inlet and outlet piping for Domestic Water Storage Tank OT-517 is required to meet PA DER regulations for public water systems and has no safety implication.

SUMMARY

No. The Fire Protection Review Report (FPRR) Subsection 4.1 and FSAR Subsection 9.2.8 were reviewed and no discussion of the double check assembly exists.

The replacement of Check Valve 0-22-035 with a spool piece actually enhances the Fire Protection System by eliminating a flow restriction to the suction supply for the automatic fire pumps. FSAR Chapter 15, "Accident Analysis" and FPRR Section 6.0, "Fire Hazards Analysis" have been reviewed.

No. This modification will enable the fire protection water supply from the Clarified Water Storage Tank OT-523 to perform its intended function as one of three suction supply sources to the fire pumps. The modification meets all design requirements of this Fire Protection Water System.

No. Unit ¹1 and Unit ¹2 Technical Specification 3/4.7.6 "Fire Suppression Water System" states that two fire pumps as well as two of three flow paths capable of providing suction to these fire pumps shall remain operable. The three sources of fire pump suction are the Unit ¹1 cooling tower basin, the Unit ¹2 cooling tower basin and the Clarified Water Storage Tank OT-523.

SER NO.: 92-169 CROSS

REFERENCE:

PMR ff88-3014B; Units 41 6 ff2 DESCRIPTION OF CKANCE:

Replacing components of the two Control Terminals (CTs) of the Vent Stack Radiation Monitoring System with equivalent parts to improve the capability of the system.

SUMMARY

No. The modification would not alter the capability of the radiation monitoring system to perform as functionally described in the FSAR. In fact, the upgraded system will facilitate the methods and means used by the operators to obtain information from the system. FSAR Sections 11.3.3 and 18.1.30.3.1 were reviewed.

No. The FSAR explains the functional requirements of the Vent Stack Radiation Monitoring System. Particulars descriptive of the displays, printers and other components used to satisfy the functional requirements are not reviewed in the FSAR.

It is noted that the proposed change adds no new failure mechanism which may compromise a safety-related system.

No. A review of the bases for Section 3/4.11.2.1 indicates that the intent of the monitoring system 'is to ensure compliance with the guidelines of 10CFR Part 20.

The modification is intended to improve the capability of the Vent Stack Radiation Monitoring System and thereby increase the margin of safety.

SER NO.: 92-170 CROSS

REFERENCE:

EO-100/200-112; Units ¹1 5 ¹2 DESCRIPTION OF CHANCE:

Changes to the subject EO procedure that incorporate information from Rev. 4 of the BWROG EPG.

SUMMARY

No. EO-100/200-112 identifies the ADS SRV's as the preferred means for accomplishing the RPV depressurization. In the event that these ADS SRV's do not function, other SRV's may be substituted in their place. In any case, the specified number of SRV's to be opened is equal to the number of ADS SRV's (6). As a result, the depressurization transient imposed upon the RPV will be within the plant design basis. Table 3.9-15 of the SSES FSAR shows that the plant design bases includes an RPV depressurization transient initiated by the opening of six SRV's which results in cooldown rates greater than 100'F per hour. When fewer than six SRV's are opened, the depressurization rate is less severe. In addition, the same FSAR Table also shows that the plant is designed to withstand even more severe depressurization transients such as those associated with a large LOCA. This large LOCA produces a transient which is more severe for the nuclear fuel, the RPV, and for the primary containment, than the transient which results from the RPV depressurization imposed through the implementation of EO-100/200-112. The large LOCA has been analyzed in the SSES FSAR.

No. The result of the rapid RPV depressurization transient which will occur through the implementation of EO-100/200-112 is a decrease in Reactor Coolant Temperature and a decrease in Reactor Coolant Inventory, both of which have already been analyzed in the SSES FSAR. The transient introduced by the large LOCA which has already been analyzed in the FSAR, is more severe than the transient initiated through the use of EO-100/200-112, Rev. 4.

No. The proposed action allows the rapid depressurization of the RPV in a manner which will be no more severe than the depressurization transient associated with an automatic ADS actuation. The existing SSES Technical Specification (Section 3.5.1.D) basis accepts this level of severity and in fact Technical Specifications require that ADS be operable during normal plant operation.

SER NO.: 92-171 CROSS

REFERENCE:

EO-100/200-104; Units ¹1 & ¹2 DESCRIPTION OF CHANCE:

Revisions to EO-100/200-104 that incorporates BWROG EPG, Rev. 4 guidance.

SUMMARY

No. The actions specified in this procedure serve to preserve the function of the Secondary Containment which is relied upon in the FSAR safety analyses for the control of radioactivity releases. The use of ADS is consistent with the use of ADS as described in Section 7.3.1.A.1.4 of the FSAR which acknowledges the manual operation of the ADS system. It allows the operator to control ADS manually instead of an automatic, when, based on an assessment of other plant conditions, it seems prudent to do so. The plant transient which results from an ADS actuation has been analyzed in the FSAR, and is less severe than other analyzed transients such as one large break LOCA transient (FSAR Table 3.9-15). The actions specified in this procedure provide guidance for restoring RB HVAC when radioactivity releases are below Technical Specification limits.

No. The manual control of ADS will introduce a plant transient which is less severe than the large break LOCA analyzed in Section 15.7 of the FSAR. (See FSAR Table 3.9-15). The actions specified to perform a manual reactor shutdown or scram also introduce transients which are within the bounds of FSAR Chapter 15 analyses. The actions specifying the use of HVAC and sump pump systems, and the verification of the proper operation of these HVAC and sump pump systems operate these systems as they were intended to be used.

m. No. The actions specified in the proposed version of EO-100/200-104 allow compliance with existing SSES Technical Specifications in that the action statements associated with Technical Specification 3/4.3.2 (Isolation Instrumentation) and 3/4.11.2 (Radioactivity Releases) are complied with. The specified actions use plant equipment as it was intended to be use and as FSAR analyses have assumed. The transients introduced by the proposed use of plant equipment to initiate a reactor scram or RPV depressurization are bounded by existing FSAR Chapter 15 Accident Analysis for LOCA and scram-producing events.

SER NO.: 92-172 CROSS

REFERENCE:

EO-100/200-114; Units ¹1 & ¹2 DESCRIPTION OF CHANGE:

The RPV Flooding Procedure (EO-100/200-114) gives instructions to maintain adequate core cooling and restore on-scale RPV water level indication in situations where RPV level cannot be determined.

SUMMARY

No. The proposed action provides steps which are carried out after the accident has already occurred. FSAR Section 6.2.1.1.3.3 was reviewed in the evaluation.

No. The purpose of this procedure is to ensure adequate core cooling when level indication is not available. It provides guidance for ensuring adequate core cooling so the proposed actions would not cause the plant condition to progress into some previously unanalyzed accident condition.

No. The RPV Flooding Basis states that adequate core cooling is assured when the active core is covered with liquid or 2-phase mixture and the conditions prevent termination of injection before the RPV is flooded above the top active fuel.

Technical Specification 2.1.4 states that RPV water level shall be maintained above top of active fuel. This Technical Specification applies to the situation where the reactor is shutdown and only heat is generated. EO-100/200-114 does not specify any operator actions which would lead to violation of this Technical Specification whenever water level indication is available.

SER NO.: 92-173 CROSS

REFERENCE:

DCP 92-9020; Unit Common DESCRIPTION OF CHANGE:

To install a new undervoltage protection scheme on the Diesel Generator E MCC OB565 bus. A new one is needed since the present one on the Diesel Generator E MCC 08565 bus does not transfer the MCC from the offsite power source to the Diesel Generator E power source or sustained degraded bus voltage below 90%

and above 30%.

SUMMARY

No. The addition of a new undervoltage protection scheme to the Diesel Generator E MCC OB565 bus does represent a change in probability of occurrence of a malfunction of equipment due to the additional relays and contacts in the existing undervoltage protection scheme. However, based on engineering judgement, the increase in probability is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculation and does not constitute a significant increase. Chapters 6 and 15 of the FSAR were reviewed.

No. The setting and verification of the setpoints of the relays and the time relays is performed in accordance with manufacturer's instruction and existing approved procedures, The random single failure during a LOCA or a LOCA/LOOP has the same impact as a malfunction with existing automatic bus transfer logic for MCC OB565. FSAR Chapters 6 and 15 were reviewed.

No. Technical Specification Tables 3.3.3-1, 3.3.3-2, and 3.3.3-3 form the bases to ensure the effectiveness of the instrumentation used to initiate the actions. The proposed action adds to Technical Specification Section 3/4.3.3 the required number of operable channels and the conditions for operability (Table 3.3.3-1), the setpoints (Table 3.3.3-2) and the response times (Table 3.3.3-3) of the new undervoltage protection scheme on the MCC OB565 bus.

SER NO.: 92-174 CROSS

REFERENCE:

DCP 92-5004; Units ¹1 K ¹2 DESCRIPTION OF CHANCE:

The following Safety Parameter Display System (SPDS) displays will be replaced: SP Load Limit, SP Press Limit, DW Instrument Temperature Limit, HT Cap Temperature Limit, and DW Spray INIT. This modification is done to correspond with the changes to the EOP curves in the latest revision to SSES EOPs. The changes reflect guidance from BWR Owner's Group Emergency Procedure Guideline (BWROG EPG) Rev. 4.

SUMMARY

No. The SPDS computers, keyboards and software are not safety-related and do not impact any FSAR analysis (

Reference:

Safety Analysis Report for SPDS - Enclosure to PLA-1882).

No. The SPDS computers, keyboards, and software are not safety-related and do not impact any FSAR analysis (

Reference:

Safety Analysis Report for SPDS - Enclosure to PLA-1882).

The failure modes after this modification remain the same, for the same systems, as that before the modification. The modification will in no way affect the availability of SPDS.

No. SPDS computer software, computer hardware and keyboards perform no safety functions and are isolated from any equipment which performs a safety-related function.

The SPDS EPG displays will be used by the Shift Technical Advisor (STA) and other control room Operations personnel as a quick indication of the reactor state in relation to the various EOP limits, as well as trends of reactor parameters either toward or away from these limits. However, SPDS will not replace the use of the actual EOP procedures, curves and charts, because the proposed hardware and software changes to SPDS do not affect any systems or functions addressed in the Technical Specifications.

SER NO.: 92-175 CROSS

REFERENCE:

EO-100/200-115, Rev. 0; Units 01 5 42 DESCRIPTION OF CHANCE:

The SSES Emergency Operating Procedure is being upgraded to the BWROG EPG, Rev. 4. The Primary Containment Flooding EO-115, Rev. 0 is being issued for the first time which is the reason for the review.

SUMMARY

No. Because the sensed RPV water level never recovers above the jet pump throat, the operator will enter EO-100/200-115, Containment Flooding and will proceed to flood the containment using water sources external to the primary containment. When the containment water level reaches the bottom of the lowest recirculation system piping, 58', the operator will vent the RPV to the environment. RPV and containment venting within the design basis is only allowed if the projected doses are within 10CFR110. The shift supervisors ensure that doses are within 10CFR110 by only authorizing RPV and containment venting if the containment radiation levels are less than 3 R/hr.

No. Flooding the primary containment to 116', the level of Top of Active Fuel (TAF), will result in submerging both drywell hydrogen recombiners. With the recombiners submerged, they will be available at 1 day and cannot be used to control H, and 0, concentrations as assumed in the FSAR analysis. Flooding to this elevation, however, is only allowed if both the RPV and primary containment have been vented. Venting will also allow the operator to control the H, and 0, concentrations. Therefore, the combustible gas concentrations will remain within the design bases limits when the containment is flooded to 116'r above.

m. No. The actions proposed by the primary containment flooding procedure do not reduce the margin of safety as defined in Section 3/4.6.1.1 of the Plant Technical Specifications. Venting, as specified by the procedure, is only allowed when it has been determined that venting will not result in doses in excess of 10CFR100, for accidents which are within the design basis.

SER NO.: 92-176 CROSS

REFERENCE:

EO-100/200-103; Units ¹1 5 ¹2 i

DESCRIPTION OF CHANGE:

The Primary Containment Control Procedure is being upgraded to the BWROG EPG Revision 4.

SUMMARY

No. The actions in this procedure are taken after the incidence of the initiating event.

This procedure implements all of the operator actions in the FSAR or the DAR which are assumed to occur when demonstrating compliance of the containment design with the regulations. A review of FSAR Sections 6.2.1, 6.2.5 and Chapter 15 was done. The actions identified are all called out in the procedures and are within the design basis. Additionally, actions are called out to restore the containment parameters to the values assumed in the safety analysis. Some of these steps require that isolations and interlocks be bypassed to accomplish the step.

No. The actions directed by this procedure for conditions that are in the design basis are designed to either: implement the actions specified in the safety analyses described in Chapters 6 and 15 of the FSAR, or restore the containment process parameters to normal values. Actions which are inconsistent with the design bases analysis are only implemented when the accident has progressed beyond the design bases.

No. The operator actions proposed in EO-1'00/200-103 do not involve changing any Technical Specification Basis.

SER NO.: 92-177 CROSS

REFERENCE:

TP-139-044; Unit ¹1 DESCRIPTION OF CHANCE:

This new procedure that affects the Condensate Demineralizer 1F106A (8-G) Anion Underlay will attempt to reduce the organic sloughage of the cation heel by building a layer of pure anion resin in each vessel after a heel cleaning.

SUMMARY

No. This procedure will have negligible adverse impacts on vessel chemistry, and thus cannot increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety. The procedure should reduce aggressive ions in the coolant, thereby lowering the potential of failure of safety components in contact with the coolant.

No. The proposed action does not create a possibility for an accident or malfunction of a different type than previously evaluated. This proposed action can only slightly increase nitrates in the reactor vessel. If this increase in nitrates is also coupled with a large organic intrusion, increased MSL radiation levels could occur, but they would remain below the MSL High Radiation MSIV isolation setpoint. Thus, this proposed action is bounded by an MSIV isolation. The overall effect of the action should reduce the concentration of aggressive anions in the vessel.

No. The effects of this procedure on vessel chemistry impurity levels are negligible and will not challenge the margin of safety of Vessel Chemistry Technical Specifications (T.S. 3/4.4.4 and 3/4.5.5), nor will it have any deleterious effects on fuel cladding integrity.

SER NO.: 92-178 CROSS

REFERENCE:

SCP E92-2052; Unit ¹2 1

DESCRIPTION OF CHANGE:

The tap settings are increased on the Unit ¹2 Generator Step Up (GSU) transformers from 525000-23600V to 537500-23600V, and at the same time, the tap settings are decreased on the Unit ¹2 Auxiliary Transformer T-12 from 23000-13800V to 22425-13800V. This is being done to indirectly reduce the operating voltage of the Unit ¹2 generator and to reduce the excitation voltage on Unit ¹2 GSU transformers.

SUMMARY

No. FSAR Section 15.2 analyzes the increase in reactor pressure from the loss of an auxiliary transformer (Table 15.2-12 and Figure 15.2-7) and the loss of grid connections (Table 15.2-1 3 and Figure 15.2-8). FSAR Section 15A analyzes the plant operation for the loss of an auxiliary transformer (Table 15A.6-28) and the loss of grid connections (Table 15A.6-29).

The proposed action lowers voltage at the Unit ¹2 main generator and reduces excitation of the GSU transformers. This does not change the probability of failure of these components.

The proposed action increases the voltage at the Unit ¹2 13.8 KV auxiliary bus and increases the excitation of the T-1 2 transformer by approximately 0.5%. The voltages of all components shall be within the limits specified in ANSI/IEEE C57.12.00-1987 and ANSI/IEEE C57.116-1989. The increase in failure probability from increases voltages by 0.5% is insignificant.

No. The most extreme case of a failure involving the BOP system is the loss of an auxiliary transformer. The most extreme case of a failure involving the main generator or GSU transformer is the complete loss of grid connections (load rejection). These cases are analyzed in the FSAR, Sections 15.2 and 15.A.

No. Unit ¹2 Technical Specification, Amendment No. 71, Table 3.3.3-1 and Table 3.3.3-2, give 4.16 KV, Class 1E undervoltage relay setpoints. The basis of these setpoints is the analysis presented in PLA-3452 (Dockets 50-387 and 50-388). This analysis is partially based on voltages at the 13.8 KV startup buses.

When a Unit ¹2 generator synch. breaker lockout occurs, auxiliary loads are transferred to Startup Bus 20. This fast transfer relies on the fact that the voltages at the unit auxiliary buses are approximately equal to the voltages at the startup bus at the time that the load transfer takes place. Voltages at the startup buses are regulated in a control band from 14025V to 14320V. Per E-AAA-695, Rev. 0, voltage at the Unit ¹2 auxiliary bus is approximately 13890V, and the proposed action will increase voltage at the Unit ¹2 auxiliary to approximately 13950V. This decreases the difference in voltage between the Unit ¹2 auxiliary bus and the startup bus; therefore, the margin of safety for the fast transfer of auxiliary loads is not reduced.

SER NO.: 92-179 CROSS

REFERENCE:

TP-069-040, 041; Unit ff1 r

DESCRIPTION OF CHANGE:

Reduce the organic contaminants in LRW collection/surge tanks. Decontaminate the LRW collection/surge tanks and associated piping.

SUMMARY

No. FSAR Sections 15.7.3 and 2.4.13.3 are affected by the proposed action. TP-069-041 will free activity from system tanks, piping, etc., for removal in LRW filters or chemical waste processing system. TP-069-040 does not increase the radioactive inventory in the radwaste system and measured liquid activities will likely not be higher than normal since thorough tank cleaning per TP-069-041 will be performed as a prerequisite to this TP.

No. The only failure that could be postulated as a result of this test procedure would be an integrity failure resulting in draining radioactive material contained in any of the affected components for TP-069-040 or TP-069-041. Any integrity failure in any of the affected components for TP-069-040 or TP-069-041 could occur now as a result of other system failures not related to this test. Since the proposed tests do not significantly affect the radioactivity contained in any LRW system 'c'omponent compared to design basis accident analysis of the concentrates waste tank, a failure as a result of this test would be no different than any possible failure. FSAR Section 15.7.3 involving failure of the evaporator concentrates tank, bounds all LRW integrity failures. The test procedures control the amount of hydrogen peroxide and catalase that can be injected to a LRW collection/surge tank set, processed through the LRW filters and demineralizer or the chemical processing system, transferred from a LRW sample tank set to a CST, or discharged from a LRW sample tank set or evaporator distillate sample tank to the river.

Based on the test controls, minimal effects on plant equipment are expected as a result of injection of the peroxide and catalase.

No. No peroxide treated water from TP-069-041 will be returned to the CST since all water will be discharged to the river through chemical waste. For this reason, the TP will not affect reactor coolant system chemistry. The reduction in organic material in the liquid radwaste stream is intended to improve reactor coolant system chemistry through reduction in contaminants returned to the hotwell (CST).

Technical Specification Sections 3/4.3.7, 3/4.4.4, 3/4.11, 3/4.12 were reviewed.

SER NO.: 92-180 CROSS

REFERENCE:

DCP 91-3003; Unit P1

/

DESCRIPTION OF CHANGE:

The new design/configuration is to restore the plant to its original pre DCP 83-019 installed configuration with parallel tank operation; however, some equipment originally installed will be left in place and spared. This is being done with the philosophy of removing unused equipment and minimizing personnel exposure.

SUMMARY

No. The presence of this spared/abandoned material in the plant which has been installed in accordance with approved procedures does not affect the operation of the Chemical Waste Neutralization tanks or any other system nor does its physical presence establish any significant hazard to the plant or its operation.

No. Sparing of non-Q conduits installed in accordance with PP&L Specification C-1035 and abandonment of non-Q instrument tubing installed in accordance with Bechtel Drawing ZJ-G16 will not cause any Safety Impact Items in the Turbine Building.

The sealing of penetrations in the Turbine Building in accordance with PP&L Specification C-1027 eliminates any potential loss of radiation shielding.

No. There is no Technical Specification section applicable to the Chemical Waste Neutralization tanks, nor are there any routine surveillance inspection or testing of system components.

SER NO.: 92-181 CROSS

REFERENCE:

DCP 92-9085; Unit 41 DESCRIPTION OF CHANGE:

To determinate and spare the power cable that feeds the space heaters for the 20 MOV's in the following systems: Control Structure Chilled Water, Feedwater, Residual Heat Removal, Reactor Core Isolation Cooling, Control Rod Hydraulics, High Pressure Coolant Injection, Reactor Water Clean-up and Reactor Vessel and Auxiliaries. This is done because MOV's are primary source of heat that could result in cable damage.

SUMMARY

No. An engineering study has determined that the electrical operability of the MOVs is not affected if the space heaters are deenergized. Therefore, deenergizing the heaters does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment. Based on the damage that occurred as a result of the space heater being energized, this modification results in a decrease in the probability of occurrence of malfunctioning of equipment.

The removal of the space heaters from service has no affect on any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR or NUREG-0776.

No. The determination and sparing of the power feed cables is in accordance with existing approved installation procedures. Electrical separation is maintained in accordance with PP&L Specification E-1012 and FSAR Section 8.3.1.11.4. This change is in compliance with Appendix R. The deletion of the space heaters in the motor and limit switch compartment was evaluated and found acceptable. Chapters 6 and 15 of the FSAR were reviewed.

No. Removal of the space heaters from service for the MOVs does not have any adverse affect on the operability, surveillance requirements or any existing margin of safety as defined in the basis in any Technical Specification applicable to the MOVs.

SER NO.: 92-182 CROSS

REFERENCE:

SCP ¹E92-1061; Unit ¹1 DESCRIPTION OF CHANGE:

Changes Unit ¹1 RPS alternate feed breakers ground fault relay setting from 10A to 15A, and formally establish the ground fault relay settings performed by Bypasses 1-92-040 and 1-92-041. This is done to make Unit ¹1 scheme identical to Unit ¹2.

SUMMARY

No. A bolted ground fault in the 480V AC primary side of the ISO-REC transformer will cause undervoltage in the 120V AC RPS power supply. This undervoltage is sensed by the EPA breakers which trip open as designed to isolate the faulted section of the electric circuit from the RPS bus, which is manually transferred to the primary power supply.

The change in the ground fault relay setting does not change the configuration of the RPS power supply system. FSAR Sections 7.2, 8.3.1.6, 15.2.6, and 7.2.2.1.2.2.14 were reviewed.

No. The nonavailability of the RPS power supply has been evaluated in FSAR Section 7.2.2.1.2.2.14. RPS's a fail safe system.

The 480V AC power supply in a ground fault situation and its effect on the RPS 120V AC power supply is investigated.

Electrical fault conditions in the 120V AC RPS power supply are sensed by the EPA breakers. Faulted power supply is isolated by the EPA breakers.

III. No. The ground fault relay setting is not specified in the Basis section of the Technical Specifications. Only overvoltage, undervoltage and underfrequency setpoints of the EPA breakers are specified in the Technical Specification.

The ground fault relay setting does not involve any change in the EPA breakers setting.

The established upper limit of the ground fault relay is 33% of the full load current.

The full load current of the 480V AC primary side of the ISO REG transformer is 52A. The ground fault relay setting of 15A is 29% of the full load current.

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