ML20198E393

Forwards 10CFR50.59 Annual Rept & Rev 5 to Updated FSAR for Seabrook Station

ML20198E393
Person / Time
Site:	Seabrook
Issue date:	12/24/1997
From:	Drawbridge B NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20198E401	List: ML20198E393
References
NYN-97125, NUDOCS 9801090090
Download: ML20198E393 (86)

Text

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North North Atlantic Energy Service Corporation P.O. Ilox 300 Yg AilanilC Seabrook, NII 03874 h

(603) 474-9721

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The Northeast Utilitiu System December 24,1996 Docket No. 50-443 NYN 97125 AR#9700125

- U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Seabrook Station 10CFR50.59 Annual Report and Revision 5 to the Seabrook Station Undated Final Safety Analysis Renort North Atlantic Energy Service Corporation (North Atlantic) encloses herein the 10CFR50.59 Annual Report and Revision 5 to the Seabrook Station Updated Final Safety Analysis Report (UFSAR). The 10CFR50.59 Report and t'ie UFSAR are submitted pursuant to the requirements of 10CFR50.59(b)(2)Tind 10CFR50.71(e), and in compliance with the 18 month schedule established in. North Atlantic's letter' and the updated schedule contained in North Atlantic's subsequent lette/. This report covers the period from January 1,1996 through June 26,1997.

UFSAR Revision 5 incorporates approved and implemented design changes and UFSAR changes identified through June 26,1997. The incorporated changes to the UFSAR have been subject to rdview in accordance with 10CFR50.59. The reviews determined that these enanges did not constitute unreviewed safety questions.

Each UFSAR replacement page includes the revision number and a vertical line (revision bar) in the left margin, adjacent to the actual change. Please note that some replacement pages have a revision nu.nber at the top of the page but no revision bar. This indicates that text position has changed as a result of text changes on previous pages and no other changes have occurred.

Summaries of the 10CFR50.59 safety evaluations for the changes incorporated in Revi.; ion 5 of

' North Atlantic letterNYN.93c80 dated May 26,1993,"10CFR59.59 Annual Report and Revision 2 to the Seabrook Station Updated Final Safety Analysis Report," T. C. Feigenbaum to USNRC.

.2 North Atlantic letter NYN-94096 dated August 26,1994," Revised Scheat e for the 10CFR50.59 Report and UFSAit Reviz. ion Submittal," T. C. Feigenbaum to USNRC.

9301090090 961224 PDR ADOCK 05000443 m-as

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'U.S. Nuclear Regulitory Commission"

NYN 97125/Page 2'

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N 4 the UFSAR are attached as Enclosure 1. Enclosure 2 provides a sui 5hry of the _10CFR50.59 l safety evaluations perfonned to support facility and procedure changes during the reportmg; period _that_ were determined to be reportable pursuant to _10CFR50.59(b)(2). Enclosure 3 is a -

listing of 10CFR50.59' safety evaluations performed during this reporting period that;were.

performed-for ' conservatism and did not meet the reporting-requirements as outlined Lin-10CFR50.59.-

This'UFSAR submittal, specifically,' Chapter 17 and Appendix 17A, addresses the reporting requirements of 10CFR50.54(a).

The. signed _ original and ten copies of the UFSAR revision'are being submitted to the Document Control Desk, Washi_ngton,-DC, along with a copy of the Regional Office, King'of Prussia, PA and a copy to the Resident inspecto. at Seabrook Stadon. This distribution complies with the frequirements of 10CFR50.4(b)(6).

Should you have any questions regarding this letter, please contact Mr. James M. Peschel,-

- Regulatory Compliance Manager, at (603) 773-7194.

Very truly yours, NORT-LANTIC E IRGY SERVICE CORP,

/d 2

BrucefDrawbridge Director - S

' ices cc:

10 J. Miller, Region 1 Administrator C.W. Smith, NRC Senior Project Manager, Seabrook Station R. K. Lorson, Senior Resident Inspector, Seabrook Station 4

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. Date: December 24,- 1997 Then personally appeared before me, the above-named, being duly sworn, did state that he is and

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.is a duly authorized officer of North Atlantic Energy Service Corporation and is authorized to execute and file the foregoing information in the name and on behalf of North Atlantic Energy

. Service Corporation and that the statements therein are true to the best of his knowledge and --

- belief.

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%W m]ov/n Susan J. McM December 24,1997 Notary Public My commission expires December 22,1998 i-t

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i ENCLOSURE 1 TO NYN-97125 Summaries of the 10CFR50.59 Safety Ev*.iuations for the Revision 5 of the UFSAR

i

. DESIGN CHANGE RECORDS Design changes documented in the following Design Change Records (DCR) were installe'd during the period covered by this 10CFR50.59 Report. A safety evaluation was performed for each DCR.

Each safety evaluation concluded that the DCR did not involve an unreviewed safety question. Summaries of these DCRs and their safety evaluations are provioed below.

DESIGN CHANGE RECORD (DCR) NUMBER:

86 298 TITLE: Turbine Generator Relaying Modification

SUMMARY

DESCRIPTION AND PURPOSE: This DCR modified the main generator sequential trip circuit in accordance with recommendations provided by General Electric Company, the generator manufacturer. The sequential trip circuit is used to trip the generator breaker following a turbine trip. Changes to the circuit were required to minimize unnecessarily opening the generator breaker before all sources of steam to the turbine have been closed off. These changes include the addition of a reverse power relay permirsive and main steam control valve position switches in the sequential trip circuit. The pwpose of these changes is to reduce the risk of damage from turbine generator overspeed during non-electrical turbine trips where the turbine rray already be in an upset condition.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support this' DCR. Although the specific dotails of the turbine generator sequential trip circuit is are not described in the UFSAR, changes to this circuit affect operation of the generator, generator breaker and electrical distribution system that are described in the UFSAR. This DCR adds a description of the revised sequential trip scheme in Section 10.2.

This DCR also corrects the description of the generator breaker tripping in UFSAR Section 8.2. Thus this DCR makes changes in the f acility as described in the UFSAR.

The safety evaluation concluded that the changes made by this DCR are consistent with the existing design criteria for the affected systems and equipment. The reliability of the electrical distribution system is maintained.

The changes are in accordance with the recommendations of the turbine generator manufacturer. The effect of the changes is to reduce the risk of turbine generator overspeed.

These changes do not affect or change the safety function of any systems or components that are credited in the UFSAR. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 96-050 1 of 32

(DCR) DESIGN CHANGE RECORD (DCR) NUMBER:

87-055 TITLE: Boron Concentration Measurement System (BCMS) Abandonment

SUMMARY

DESCRIPTION AND PURPOSE:

The original description of the DCR as presented and approved in 1987 was to modify BCMG (CS-TK-132) and BCMS (CS-CPA46). Since the original conception of this work it has been decided that the system does not function as required and that these modifications and the installed system is no longer required. The scope of the DCR has been revised to abandon in place the entire system.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this DCR. The safety evaluation concludsd that the ' CR made changes in the facility as described in the UFSAR (Section 7.7.1.10, J

Table 7.7 2, and Figures 7.710 through 13). The Boron Concentration Measurement System is designed for use as an advisory system. It is not designed as a safeguards system or as a component of a safeguards system. The Boron Concentration Measurement System is not part of a control element or control system, nor is it designed for this use.

No credit is taken for this system in any accident analysis. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 96 001 DESIGN CHANGE RECORD (DCR) NUMBER:

91-026 TITLE:

Removal of "PEPCON" Chlorination System

SUMMARY

DESCRIPTION AND PURPOSE:

The Circulating Water System was designed to use one of two sources of continuous low-level chlorination to control the growth of marine organisms. The two sources of chlorination were the *PEPCON" generation system or from bulk solution in storage tanks.

The "PEPCON" generation system proved to be operation and maintenance intensive while the use of bulk solution in storage tanks was more effective. Phase I of this DCR removed the "PEPCON" generation system from plant design and modified the reniaining Chlorination System piping and relocated process instrumentation. Phase ll of this DCR provided design information to support the installation of a machine shop in the space created in the Chlorination Building by the removal of the "PEPCON" system.

SAFETY EVALUATION

SUMMARY

A 10CFR 50.59 evaluation was performed for this DCR. The determination of safety evaluation applicability concluded that the DCR made changes in the facility as described in the UFSAR. The modification affected UFSAR Sections 1.2.2.18 and 10.4.5.2 and Figures 10.4-5, 9.3 6 sh.1, 8.3-1,1.2-57 and 1,2 58. This modification affected only non-safety related portions the Chlorination System, Service Air System, Potable Water System, Fire Protection System, and Non 1E Electrical Distribution System.

The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 91-035 2 of 32

4 DESIGN CHANGE RECORD (DCR) NUMBER:

92-059 TITLE: ~ Hotwell Level Control System Replacement -

SUMMARY

DESCRIPTION AND PURPOSE:

A replacement Hotwell Level Control System was installed in order to reduce the required system maintenance, minimize the. amount of testing for air inleakage, and improve the functional reliability of the control system. The new controls will operate the spill / makeup valve in a full'open/close fashion, and the rance will be increased to ensure that the process remains onscale during any plant transient. This DCR also installed test thermowells/ connections on "A" hotwell.

SAFETY EVtNJATION

SUMMARY

A 10CFh50.59 cafety evaluation was performed for this DCR. The determination of safety evaluation applicability concluded that the DCR made changes to the facility as described in the UFSAR (Section 10.4.1, Figures 10.4 7 sh.1 and 5, Figure 1.2-37, Chapter 15.2.7).

Design and safety evaluations dependent on these provisions remain valid, no acceptance levels or system limitations have been altered and no design or safety margins have been

. reduced. The safety evaluation concluded that no unreviewr - fety question was created by this DCR.

UFCR 93-051 DESIGN CHANGE RECORD (DCR) NUMBER:

93-007 TITLE: Leased Makeup Water Treatment System

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR made provisions for replacing the existing Water Treatment (WT) System with a leased Makeup Water Treatment System from IONICS Inc. The DCR also made the ultraviolet sterihzation unit installed under TMOD 90-0047 a permanent installation, as well as providing the details for removing existing equipment that will not be required when the new system is installed.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evalcation was performed for this DCR. The determination of safety evaluadon applicability concluded that the DCR made changes to the facility as described in

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the UFSAR (Sections 1.2, 8.3, 9.2.3, 9.4.12, 9.5, 9.5.1.2.c.5, 10.2.2.2, 11.3.2 also

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Appendix "A", and Figures 8.31 and 10.2-1). Design and safety evaluations dependent upon these provisions remain valid, no acceptance levels or system limitations have been altered and no salon or safety margins reduced. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 93-054 L

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. DESIGN CHANGE RECORD (DCR) NUMBER:

_94-003 TITLE: Distributed Control System

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR installed a Distributed Control _ System (DCS). The existing FW heater controls, local and-pneumatic, were replaced.

New instrumentation (level)- connections were-

- installed. The broken sightglasses and level switches were replaced. Also, the logic for CO V75 (21/22 FW heater b) pass) was modified such that the bypass valve may not be closed unless all the inlet and outlet isolation valves are fully open.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed.

The determination of the safety evaluation applicability concluded that the DCR did make changes to the facility as described in the UFSAR (Section 10.4.7, and Figures 1.232,1.238). The functions and

- failure modes remain bounded by the current safety analyses.

The DCR !a an enhancement and did not adversely affect the design basis or function of safety related structures, systems, or components. The safety evaluation concluded that the DCR did not involve an unreviewed safety question.

UFCR 96-059 DESIGN CHANGE RECORD (DCR) NUMBER:

94-039

- TITLE: Feedwater System Overpressure Mitigation

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR provided changes to valve (FWlV's, FWRV's) stroke times, FW pump speed, and FW isolation to respond to conditions exhibited by the Main Feedwater System (FW) after a Unit trip. The automatic FW isolation results in an overpressure condition in the FW system that briefly exceeds the nominal system design pressure. The overpressure 3

condition is caused by the combined sum of the Total Displaced Head of the Conder' sate, Heater Drain and FW Pumps under low flow conditions, and the rapid hydraulic tran.iient from the FW isolation.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this DCR. The determination of the safety evaluation concluded that the DCR made changes in the facility as described in the UFSAR (Section 6.2.1, Tables 6.2 13, 6.2-14, 6.2 15, 6.2-66, 6.2-68, 6.2-69a, 6.2-69b, 6.2-83, 7.3 2, and Figures 7.77,7.22,-7.71,7.2-1,7.213 sheets 1 and 2 and Table 16.3-2 [all included in UFCR 95-053] and Chapter 15 [UFCR 94-056]). Only balance of plant control systems are affected. Design and safety evaluations dependent upon these provisions remain valid, no acceptance levels or system limitations have been altered and no design or safety margins have been reduced. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 95-053 t

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DESIGN CHANGE RECORD (DCR) NUMBER:

94-046 TITLE: CS Filter CS-F-4A and CS-F-48 On Line Differential Pressure indication

SUMMARY

DESCRIPTION AND PURPOSE:

' This DCR replaced Barton differential pressure indicating switches with Rosemount differential pressure transmitters to allow continuos on-line monitoring of the pressure drop across the two Reactor Coolant Pump (RCP) Seal Water Injection filters CS-F-4A and CS F-

48. The capability to obtain and trend filter differential pressure values while operating at power aids in the evaluation of potential RCP seal degradation.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was prepared to support this DCR. This DCR resulted in changes in the facility described in UFSAR Section 9.3.4.5.m and Figure 9.3-29. Replacement of the existing pressure differential indication switches _with pressure differential transmitters having the same span, and that provide indication in the control room does not change the function, scope or intent of the RCP Seal injection System. Providing differential pressure indica..sn in the control room for the RCP seal injection filters will aid in the evaluation of potential RCP seal degradation.

The differential pressure indication is non safety related and does not interact with any safety related system. The transmitter power supplies do not add any new Class 1E or non Class 1E interfaces. The effects of a failure are the same as the original design. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 96-021 DESIGN CHANGE RECORD (DCR) NUMBER:

94-049 TITLE: Chlorination System Modifications (CA-0,2,4,5)

SUMMARY

DESCRIPTWN AND PURPOSE:

This design coordination report provided modifications that increase the reliability of the Chlorination System. The changes include the incorporation of an additional chemical injection pump, adding flexible connectors at the pumps' suction and disch. ge nozzles, installation of isolation valves on the pumps' safety velves and the replacement of the pumps' suction strainers.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 se'ety evaluation was performed for this DCR. The determination of safety evaluation applicability concluded that the DCR made changes to the f acility as described in the UFSAR (Section 10.4.5, Figure 10.4-5). The safety evaluation concludad that the DCR was an enhancement and did not adversely affect.the design basis or function of safety related structures, systems, or components. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 95 008 5 of 32 L

1 DESIGN CHANGE RECORD (DCR) NUMBER:

95-011 TITLF.: Fish Counting House improvements

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR provided improvements to the Fish Counting facilities in the Circulating Water (CW) Pumphouse. _ Improvements completed per this DCR included installing additional

- ventilation and lighting, providing higher pressure water to clean the fish pit, providing a motorized trolley and hoist on thc existing monorail and providing a salt water flush lins for bivalve plankton not cleaning.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed for this DCR.

The CW building air handling' system is shown on UFSAR Figure 9.4 22. The fish counting facility is shown on Figure 1.2-46 and Figuru 10.4-4. _The changes made by this DCR required changes to those drawings thus this DCR makes changes to the facility as described in the UFSAR.

The drawing changes are necessary because of the changes to the ventilation equipment and the additional fiieloading created as a result of the additional fiberglass and PVC piping added by the DCR. The safety evaluation concluded that the changes do not alter the design, function or operation of systems in the CW Pumphouse. The changes do not compromise the security of the structure and there is no impact on the on systems governed by Technical Specifications, Appendix B to the Operating License Environmental l

Protection Plan or to the Appendix R Report. The Appendix A report is updated as a result of the changes made by this DCR.

The changes did not affect any safety related equipment or-structures. The safety evaluation concluded that.no unreviewed safety question was created by this DCR.

UFCR 96-007 DESIGN CHANGE RECORD (DCR) NUMBER:

95-014 TITLE: Halon System Removal

SUMMARY

DESCRIPTION AND PURPOSE:

In September 1987, the Montreal Protocol was signed to eliminate the use of substances that deplete the Ozone Layer. The production of Halon ceased on January 1,1994. This DCR was part of the Seabrook Station action plan to reduce the use of Halon 1301. The DCR removed the Halon system in the Secondary Alarm Station (SAS) Computer Room and the Halon system in the Administration Building Document Control Center (DCC). The SAS Computer Room is an unoccupied masonry block room, approximately 15'x15' with a concrete floor and ceiling located at the 75' elevation in the Turbine Building.

The Administration Building DCC Room was designed and installed as. a storage area for construction records. The construction records have since been moved to another storage area.

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The DCC Room is separated from the Turbine Building by-a three hour rated block wall.

_The-_ fire detection control panel for the SAS Computer Room and the Administration Building DCC was rewired to delete the Halon system release function. The fire detection panel remains active and alarms in the Control Room.

SAFETY EVALUATION

SUMMARY

- A 10CFR50.59 evaluation was performed for this DCR.

The determination of safety _

evaluation applicability concluded that the DCR made changes in the facility as described in the UFSAR.

The modification affected UFSAR Section 9.5.1 and Section B, " Fire Protection System Description" in "Seabrook Station Fire Protection Program, Evaluation and Comparison to BTP 9.51, Appendix A."

The elimination of the Halon systems in the SAS and the Administration Building DCC will not significantly reduce the fire protection capabilities in these areas. The SAS Computer Room and the Administration Building DCC areas are surrounded by concrete and have active fire detection systems. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 96-002 DESIGN CHANGE RECORD (DCR) NUMBER:

95-022 TITLE: Full Voltage Light Bulb Replacement

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR corrected concerns where incandescent light bulbs when shorted could operate the circuit protective device and disable a safety related circuit. A review of safety related circuits a+. Seabrook Station identified three applications that use full voltage light bulbs:

The Solid State Protection System (SSPS), Radiation Monitors (RM), and status monitor light for Containment isolation valve position. For SSPS and status monitor lights the bulbs were replaced by LED's; for the RM the bulbs were removed.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed for this DCR. The determination of the safety evaluation applicability concluded that the DCR mcde changes to the facility as described in the UFSAR (Chapters 7 and 8, Figures 12.30-19&20, 8.3 10). The safety evaluation concluded that the DCR is an enhancement and did not adversely affect the design baris or function of cafety related structures, systems, or components. These changes have been designed and installed to existing criteria. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 95-045 7 of 32 i

DESIGN CHANGE RECORD (DCR) NUMBER:

95-035 TITLE:

13.8 KV to 480 Volt Unit Substation Transformer Replacements

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR evaluated and approved replacement 13.8 KV 480 Volt transformers for the 480 Volt unit substations. The unit substation transformers affected by this DCR are classified as nonsafety related. Although classified as nonsafety related, transformers located within the nuclear island are required to be seismically supported to ensure that their collapse or failure does not affect the operation of any nuclear sI,tv related equipment.

The replacement transformers are of an improved design that complies with design basis requirements, are compatible with the existing unit substations, and will not affect the functionalintent of the electrical distribution system.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was prepared to support this DCR. This DCR makes changes in the facility as described in the UFSAR. The 13.8 KV 480 Volt distribution system is described in Chapter 8 of the UFSAR. The replacement transformers are of an improved design that will enhance the reliability of the nonsafety-related electrical distribution system. The changes made by this DCR are consistent with the design criteria for the affected systenis and equipment.

The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 95-059 DESIGN CHANGE RECORD (DCR) NUMBER:

95-043 TITLE: CAS-P-252/254 Pump Controller Modifications

SUMMARY

DESCRIPTION AND PURPOSE:

Changes in the way the secondary plant chemistry is maintained required changes to the secondary Chemical Addition System (CAS) pump controllers.

The original pump controllers were designed to operate in automatic and manual mode.

The chemistry department has had stability problems operating the controllers in the manual mode, and automatic control of the controllers is no longer necessary.

The manufacturer of the-existing _ controllers was contacted and they confirmed the existing controllers are not very effective in the manual mode.

This DCR replaced the original controllers with new controllers that are more reliable when operating in the manual control mode.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support this DCR. The UFSAR does not explicitly describe the CAS pump controllers, however UFSAF. Figure 9.3-15 contains action blocks that suggest that automatic control of the CAS pumps is available from monitors that measure the condensate pH, hydrazine and conductivity levels. This DCR

- removed the automatic control feature of the system and provided for only manual control of the stroke of the CAS pumps from the new controllers. Thus this change is considered

- a change to the facility as described in the UFSAR.

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The safety evaluation concluded that the CAS system is a non safety system located in the non seismic Turbine Building and the controllers do not provide any type of safety function.

A failure of the new controllerr will not cause a subsequent failure of any safety related-equipment.

The new controllers are more reliable in the manual mode than the old controllers that they replaced. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 95-056 DESIGN CHANGE RECORD (DCR) NUMBER:

95-047 TITLE: Amp-C p and Circuit Breaker Instantaneous Trip Time Test Se'noint

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR revised the instantaneous trip response time acceptance criteria for Molded Case Circuit Breakers (MCCB's) and Amp-Cap Motor Circuit Protectors (MCP's) based on recommendations provided in current industry standards. MCCB's and MCP's affected by this DCR are used both in safety related and non safety related applications. Use of these revised acceptance criteria is in accordance with applicable industry standards and will continue to ensure that the protective devices are functioning properly.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this DCR. The determination of the safety evaluation applicability concluded that the DCP made changes to the f acility as described in the UFSAR (Chapters 3,16, Tables 16.3-8, and 16.310). The changes made by this DCR increase the maximum allowable instantaneous trip response time that is used during field verification testing. This time is selected based on applicable industry standards referred by the manufacturer. The changes have been evaluated within design basis calculations with no impact on results or conclusions. These changes do not introduce any new failure modes for the protective devices. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 95-064 DESIGN CHANGE RECORD (DCR) NUMBER:

96 06 TITLE: Pressurizer Heater Abandonment

SUMMARY

DESCRIPTION AND PURPOSE:

Heater number 8 in the Pressurizer Back Up Heater Group D is shorted and as a result, unusable..The heater was removed from service by Temporary Modification TMOD 88-19.

Heater number 8 will not be replaced and this DCR provided for the permanent abandoning in place of this heater.

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SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support this DCR, UFSAR Figures 5.1-4 and 8.31 and Table 5.1-1 describe the number of heaters, the capacity of Back-Up Group. D heaters and the capacity of the total capacity of the Pressurizer heaters, respectively, it was necessary to change theso figures and table as a result of implementing this DCR. The safety evaluation completed to support this DCR concluded that the Group D heators are non safety related and are not powered from an emergency bus. The Technical Specification required heaters, Back Up Groups A and B, are not affected by this DCR. The total heater capacity reduction caused by abandoning this heater is 23.08 kW,'or approx mately 1.3%. The abandonment of a single pressurizer heater does not affect the operation of any safety related system, structure or component.

The'sesfety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 97-009 DESIGN CHANGE RECORD (DCR) NUMBER:

96 013 TITLE: CAS Hydrogen Sensor Relocation

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR relocated the hydrogen detectors used to monitor Locked High Rad Areas in Containment in order to make detectors accessible for calibration during plant operation.

One detector will be used to monitor each of these areas. The remaining detectors will be removed from these areas with conduit abandoned in place.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety 1 valuation was performed for this DCR. The determination of the safety evaluation applicability concluded that the DCR made changes to the facility as described in the UFSAR (Figure 9.3 5). Design and safety evaluations dependent on these provisions remain valid, no acceptance levels or system limitations have been alterert and no design or safety margins reduced. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 96-051 V

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- DESIGN CHANGE RECORD (DCR) NUMBER:

96-015

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TITLE:

Reactor Vessel Level Instrumentation (RVLIS) Classification Correction -

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR changed the P&lD piping r,afety class designators for the Reactor Vessel Levei instrumentation System (RVLIS) and wide range Reactor Coolant System {RCS) pressure sensing lines from Non-Nuclear Safety Related (NNS) to Safety Class 2 non-ASME to comply with UFSAR commitments, added the use of commercial grade dedication as an UFSAR exception to the Seabrook compliance with Regulatory Guide 1.151 on instrumentation installation, clarified the UFSAR discussion on instrument sensing lines penetrating the containment, corrected the UFSAR discussian and loop drawings relatlng to which RCS flow tap is shared by the redundant instruments, and made other minor drawing-and UFSAR changes. There were no physical changes associated with this DCR. The minor drawing and UFSAR changes are corrections to bring the documents in agreement with the existing design and did not require a safety evaluation.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was prepared to support this DCR.- There were no physical changes associated with this DCR. The safety classification upgrade of the instrument sensing lines documents the safety classification required to meet the commitments to Regulatory Guides 1.97 and 1.151, as modified by UFCR 96-013. This upgrade is acceptable without any material change out due to the quality of the original installation and the commercial grade dedication of the isolation and vent valves.

The sensing lines were installed by United Engineeis and Constructors under their QAS 5 installation program. Within this program, material was procured as safety related and the installation design required that the sensing lines remain functional after a seismic event.

The components supplied by Westinghouse include qualified transmitters, hydraulic isolation devices, capillary tubing, and vent valves located inside the containment. The isolation and outside containment vent valves were commercial grade dedicated to meet the modified requirements of the North Atlantic commitment to Regulatory Guide 1.151; therefore, their failure rate will be the same as qualified valves.

Correcting the UFSAR and the control loop diagram to indicate that the high pressure tap, not the low pressure tap, is shared by the RCS flow transmitters brings these documents into agreement with the existing design. This is the standard Westinghouse design and is

- required so that the rupture of a sommon low pressure tap could not prevent a low flow reactor trip. A common high pressure tap rupture would result in a spurious trip. The

- safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 96-013 11 of 32

DESIGN CHANGE RECORD (DCR) NUMBER:

96-019 TITLE: RCP Seal No.1 Leakage Outside Operating L!mits

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR corrected ~ a discrepancy between actions specified in station procedu.es regarding RCP No.1 seal high leakoff flow, and actions stated in the UFSAR (Chapter 5.4, Section 5.4.1.3a) and the RCP Technical Manual (FP54502). This discrepancy is a result of information contained in Westinghouse Technical Bulletin 93-01 RO. Documentation will be changed to reflect the proper sequence of No.1 seal high flow leakoff isolation.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this DCR. The determination of safety evaluation applicability concluded that the DCR made changes to procedure as described in the UFSAR, Chapter 5.4, Section 5.4.1.3a Design and safety evaluations dependent upon these provisions remain valid, no acceptance levels or syster-limitations have been altered and no design or safety margins have been reduced. The safety evaluation corscluded that nn unreviewed safety question was created by this DCR.

UFCR 96-035 DESIGN CHANGE RECORD (DCR) NUMBER:

96-021 TITLE:

ECCS Throttle Valve Restriction Orifice Addition

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR added restricting orifices to the Emergency Core Cooling System (ECCS) hot leg and cold leg injection lines. Each of the injection lines contains throttle valves that ensure the ECCS injection flow will be balanced equally among the lines. The throttle valves also limit ECCS pump runnout during ECCS injection and recirculation phases. The restricting orifices also provide additional flow rc.tstance that allows the throttle valves to be opened further. Opening the throttle valve further ehminates the potential for debris clogging of the valves arid reduces the fluid velocity in the seat area eliminating the potential for erosion.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was prepared to support this DCR. The addition of the orifices affects UFSAR Figures 6.3-4 and 6.3 5.

The safety evaluation concluded that there is no change to the overall system resistance as a result of this change. The throttle valves havc been opened more to offset the increase in resistance provided by the orifices.

With the throttle valves open more, the potential for crosion is decreased, resulting in an increase in the long term reliability of the system. The operation of the ECCS injection pumps and flow control valves is not affected by this change. The piping changes are qualified to the same codes and standards as the original installation. Thus no unreviewed safety question was created by this change.

UFCR 90 056 12 of 32 a

4 DESIGN CHANGE RECORD (DCR) NUMBER:

96 025 TITLE: Alternate Cooling for Charging Pump I.ube Oil Coolers

SUMMARY

DESCRIPTION AND PURPOSE:

'This DCR added a cross connect from the Fire Protsetion (FP) System to the Primary Component Cooling Water (PCCW) System. This cross connect would be used to provide cooling water to the lube oil coolers of the Emergency Core Cooling System (ECCS) high head pumps, the Centrifugal Charging Pumps (CCP), when the normal cooling to the CCPs, the PCCW System, was not available. A Demineralized Water (DM) System connection to the CCP lube oil coolers was also provided by this DCR. The DM connection allows for cooling during outages when the PCCW System is not available and for flushing the tube oil coolers following use of the FP System cross connect.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was completed to support this MR The changes made bv this DCR affect the PCCW System figures in the UFSAR as dtoicted on Figures 9.2-4 shc.et 1 and 9.2-6 sheet 1; and FP System figure as depicted on Figure 9.5-8. The procedures for responding to a loss of PCCW were also affected by this DCR, therefore the facility and procedures as described in the UFSAR were affected by this DCR.

The addition of the alternate cooling systems to the CCP lube oil cooler does not affect the normal operation of the PCCW, FP or DM systems. This change increases the reliability of the Reactor Coolant Pump (RCP) seal injection function provided by the CCP, reducing the

- probability of RCP seal degradation following a loss of PCCW. This change does not affect the ability of the PCCW System or the CCPs to perform their safety function in mitigating the consequences of accidents in which they are credited.

The safety evaluation concluded that no unreviewed safety qcestion was created by this DCR.

UFCR 96-044 DESIGN CHANGE RECORD (DCR) NUMBER:

96-031 TITLE: Containment Equipment Hatch Acces Platforms

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR installed a permanent platform on either side of the equipment hatch that provided access to the hatch swing bolts. The platform allowed workers to remove and install the equipment hatch with greater safety and in less time. This DCR also relocated a fire protection hose reel to avoid interference with one of the platforms.

13 of 32

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DESIGN CHANGE RECORD (DCR) NUMBER:

96-031 (Continued)

' TITLE: Containment Equipment Hatch Access Platforms SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was prepared to support this DCR.

UFSAR Section 3.8.2.1.b describes the containment equipment hatchJ This UFSAR section was lavised to reflect the changes being made to the equipment hatch by this DCR. The safety evaluation concluded that the equipment hatch access platforms were designed and fabricated in

-accordance with the design requirements listed in the UFSAR. The platforms have been-designed so that they will not fail in such a manner as to damage equipment required for safe shutdown or decay heat removalin the event of a safe shutdown earthquake.

The platforms will not impact nearby safety related systems, structures or components during an earthquake. The design of the piping modifications complies with the design requirements for seismically supported non safety related components in a Seismic Category 1 area. The change in the piping will not impact nearby safety relater' systems, structures or components during a seismic event, The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 96 058 DESIGN CHANGE RECORD (DCR) NUMBER:

96-034 TITLE:

EFW Flow Control Valve Time Delay

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR installed changes to upgrado certain EFW circuits. A time delay relay was added to th: EFW High Flow isolation protection circuit.

This design change filters out the spurious control signals that result from valve reset operations, thus eliminating undesired flow control valve isolations. The position inputs from MS V127 (SG A isolation) and MS-V128 (SG B isolation) were removed from the EFW A TRN INOP logic, These valves are not expected to be closed more frequently than once per year and are only closed in modes where the steam driven EFW pump is not required to be operable. The position indication from MS V129 (EFW turbine overspeed trip valve) was added to the logic for the EFW A TRN INOP circuit, This valve is used to deliberately shutdown the turbine during quarterly pump testing.- Regulatory Guide 1.47 requires that automatic system level indication be provided in the Control Room to alert operators that the equipment actuated-by the protection system is inoperable when the valve is closed.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed on this DCR. The determination of safety evaluation app;icability concluded that the DCR made changes to the facility as described in the UFSAR (Sections 6.8, 7.1.2.6, 9.2.6,15.0,15.1.5, and 15.2,8, Tables 6.8 2, and-7.4-1). Design and safety evaluations dependent 'Opon these provisions after review, remain valid, no acceptance levels-or

m limitations have been reduced and no design or safety margins have been reduct. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

.UFCR 96-065' 14 of 32 4

DESIGN CHANGE RECORD (DCR) NUMBER:

96-035 TITLE: Stud Washing Jib Crane on Containment El. 25' 0"

SUMMARY

DESCRIPTION ANO PURPOSE:

This DCR provided for the installation of a one ton jib crane at Elevation 25' in Containment. The crane is for reactor stud maintenance activitas during outages. The jib -

boom and trolley are commercially purchased products. The one ton hoist is drawn from the Tool Crib when needed. -The trolley and hoist are removed from Containment prior to normal plant power operations. The jib crane mast and boom are supported by an 8" x 8" tube steel support attached to the outside of tb Bio Shield wall.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the installation of this DCR.

The DCR includes a permanent crane mast and boom that will be used with a detachtble trolley and hoist. UFSAR Figure 2,1 was revised to show the new crano. The safety evaluation completed to support this DCR detailed the construction and orientation of the crane. The entire crane assembly is NNS seismically supported and qualified for potential seismic forces. There are no safety related systems or compenents within the affected semi-circle of the crane's semi-circle limit of travel. The crane will not be used in Modes 1 through 4 and the trolley and hoist will be removed from containment during these modes.

The safety evaluation concluded that no unreviewed safety quedion was created by this DCR.

UFCR 96-060 DESIGN CHANGE RECORD (DCR) NUMBER:

97-002 TITLE: SI Test Relief Pa.

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR provided a relief path from a partially opened drain valve to the Containment equipment sump via a tubing run. This was required as the piping adjacent to the Si test line inside the containment does not have thermal relief protection (as per Generic Letter 96-06) during a design basis accident, The drain valve that will be partially opened is SI-V228. This drain valve is at an elevation that will preclude draining the entire system line during normal operation SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed for this DCR The determination of the safety evaluation concluded that the DCR did make changes to the facility as described in the UFSAR (Section 6.2, 6.3, 6.2.2.2, 6.2.4.2, 5.2.5.5, / Figures 6.2 68, 6.3-3, and 6.2-91). Design and. safety evaluations dependent upon these provisions.remdn valid, no acceptance levels or system limitations have been altered and no design or safety margins

. reduced. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 97-001 15 of 32

.~

DESIGN CHANGE RECOPD (DCR) NUMBER:

97 03 TITLE: Apperdix R Modifications

SUMMARY

DESCRtPTION AND PURPOSE:

This DCR made chranges to the Appendix R report to correct discrepancies identified during the 10CFR50.54(fs review of the Appendix R Program 'These discrepancies involved the identification of / ppendix H cables, raceways, and fire zones. The changes are consistent

.with the requirernents of Appendix R. The Primary Component Cooling Water Loop A & B Supply Header Temperature inputs to the Main Plant Computer System that are being affected by this DCR are classified as non-safety related. The plant equipment (FW V 347, SWA FN-408) and the Cable Scheduling Prngram that are affected by the documentation changes are classified as safety related.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this DCR. The determination of the safety evaluation applicability concluded that the DCR made changes to the facility as described in the UFSAR (Chapter 7, Section 7.4, and Table 7.41, Appendix R Report). Field work associated with this DCR involves the deletion of redundant inputs to the MPCS for PCCW Loop A & B Supply Header Temperatures. Other changes to the Appendix R Report are paperwork only corrections that are consistent wi'.h the requirements of Appendix R. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 97-004 DESIGN CHANGE RECORD (DCR) NUMBER:

97-004 Ti1LE: MS-V394 Control Modifications

SUMMARY

DESCRIPhOid AND PURPOSE:

This DCR addressas the issue of a Main Stoam Line Break (MSLB)/ Feedwater Line Break (FWLB) on the E-11 A Steam Generator which would result in insufficient pressure at MS-V393 to operate the steam driven pump (P-37A). The single active failure of the Train "B" SSPS would prevent automatic opening of MS-V394 to supply the steam driven feed pump from E-11B Steam Generator.

Also, this would prevent an auto start of the electric driven feed pump (P 378). This DCR and UFCR provided the design details and documentation necessary to support a modification to MS V394 to provide for a Train 'A' actuation signal.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed for this DCR. The determination of the safety evaluation applicability concluded that the DCR made changes to the facility as described in the UFSAR (Sections 6.8, 7.3, 7.4, 7.5, 7.6, 10.3.2.5, 15.0,15.1.5 and 15.2.8, also Tables - 6.2-83 and - 6.8-2, and Figure 10.3-1 ). The design and safety evaluations dependent on.these provisions remain valid, no acceptance levels or system limitations have been altered and no design or safety margins reduced. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 97-015 16 of 32 a

r 5

LDES6GN_ CHANGE RECORD (DCR) NUMBER:

- 97 005 J TITLE: Reactor Coolant Drain Tank (RCDT) Relief Path -

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR installed a relief valve (at an existing test connection) on the discharge piping l from the RCDT to provide _ thermal relief protection (as per Generic Letter 96-06) during a -

design basis accident.' The manual: Isolation was categorized _ as a I)cked open valve to ensure a flowpath is not jeopardized during an overpressure condition.

SAFETY EVALUATION

SUMMARY

- A 10CFR50.59._ evaluation was performed for this DCR. The -determination of safety

. evaluation applicability concluded that the DCR made changes to the facility as described in the UFSAR-(Sections 6.2.4, 9.3, Figures 6.2 91 sh. 7 and 9.3-17). Design and -safety evaluations dependent upon these provisions remain valid, no acceptance levels or system

-limitations have been altered and no design or safety margins have been reduced. The safety evaluation concluded that the DCR did not involve an unreviewed safety question.

UFCR-97-012 DESIGN CHANGE RECORD (DCR) NUMBER:

97 007' TITLE: Increase Stroke Time for Containment Isolation Valves, CAP V1,V2,V3,& V4

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR increased the required closing stroke time for Containment Air Purge (CAP) valves V1, V2, V3 and V4 from 3 to 5 seconds. This DCR also installed modifications to the CAP air handling system inlet ducts, adjacent to the refueling pool. The modifications 1

to the inlet ducts were necessary to maintain the validity of the existing UFSAR Chapter 15

- 4 Fuel Handling Accident Analysis.

The inlet duct modifications delay the transport time for radiological gasses released during a fuel handling accident to reach the CAP isolvion valves. The increased valve stroke time

was necessary because of difficulties in attaining the previous 3 second limit.

SAFETY EVALUATION

SUMMARY

-- A 10CFR50.59 safety evaluation was completed to support the changes made by this-DCR UFSAR Table 15.718 lists the allowable stroke time of the' CAP valves. As a result of the changes made by this DCR the table was modified. Thus.this DCR changed the

_ facility as described in the UFSAR. The safety evaluation concluded that the modifications -

to the inlet ducts would ensure that the assumptions concerning the mixing of radioactive gasses, released during n_ fuel handling accident, with the containment atmosphere prior to the release of those gasses to the environment, are valid.

i 17 of 32 1

DESIGN CHANGE RECORD (DCR) NUMBER:

97-007-(Continued TITLE: Increase Stroke Time for Containment Isolation Valves, CAP V1,V2,V3,& V4 The safety evaluation also concluded that increasing the stroke timo of_ the CAP valves from 3 to 5 seconds left ample margin with regard to the stroke time of 10 seconds credited in the original FSAR and discussed in the NRC Safety Evaluation Report. The total time for the CAP valves to close once a closure signal is generated is 6 seconds including control loop time.

This cornpares with the estimated 35 second transport time of radioactive gasses to the CAP valves. The safety evaluation concluded that adequate margin existed to prevent the direct release of radioactive gasses through the CAP system.

The safety evaluation concluded that the changes made by this DCR did not create an unreviewed safety question.

UFCR 97 024 DESIGN CHANGE RECORD (DCR) NUMBER:

97 008 TITLE: Prevention of Overpressurization in Containment Penetrations and Adjacent Piping.

SUMMARY

DESCRIPTION AND PURPOSE:

Generic letter 96-06, ' Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions," required a review of containment penetrations to withstand internal pressures that may develop due to thermal heat up of the piping in the post-LOCA containment environment. The review process looked at the containment penetrations and the adjacent piping. Several containment penetrations and adjacent piping h

required modification to fully meet the intent of Generic Letter 96 06. A reliet valve was added to the Si test line piping to prevent overpressurization. The relief valve added by DCR 97-005 was relocated to a more suitable locatinn on the RCDT discharge and the test connection was restored. Finally, several containment valves will be delineated as " Locked Open" pei Criteria 6 of the locked valve program (valves in relief path).

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 was developed to support this DCR. The changes made by this DCR were completed to ensure that the containment isolation system functions properly, to prevent the release of fission products during accident conditions and to ensure that no containment leakage occurs. The containment isolation systems, Safety injection System and Waste Liquid Drain System are all discused in the UFSAR. All three systems are also depicted in figures in the UFSAR. This DCR changes the figures and otherwise affects the systems as described in the UFSAR and thus changes to the UFSAR in Sections 6.2.4 and 3.9.6 2, Tables 3.9 (B-27), 3.9 (N-11), Figures 6.3-3, 9.1-2 sh. 3, 9.2-1/, 6.2.91 sh. 3, 5.1-4, 9.3 11, 6.2-91 sh. 7, 9.318 and 9.3-27.

The components added are seismically supported to ensure they will not interfere with any safety related equipment during a teismic event. The valves designated as " Locked Open" valves are manual valves, none of which is required to automatically operate in order to mitigate the consequences of any accident. 'The only creoible failure scenario associated 18 of 32 a

DESIGN CHANGE RECORD (DCR) NUMBER:

97-008 (Continued)

)

TITLE: Prevention of Overpressurization in Containment Penetratic,as and Adjacent Piping.

I with this DCR is the possible inadvertent lifting of a relief valve if that were to happen the discharged water would be collected in the centainment equipment sump and processed accordingly. The discharged water would not interfere with any safety system. The safety

- evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR 97-019 DESIGN CHANGE RECORD (DCR) NUMBER:

97-022 TITLE:

Polar Crane Auxiliary Holst Rerate to 53 Tons

$UMMARY DESCRIPTION AND PURPOSE:

This DCR provided the documentation for rerating the polar crane auxiliary hoist from 50 to 53 tons. The 103 foot circular span polar gantry crane was fabricated by Whiting Corp.

Whiting Corp. calculation FP57896 provides the technical justification for the rerate. The change was made to provide additional margin for the lift of the hatch cover, lifting beam, rigging and equipment hatch platforms installed per DCR 96-0031.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed for this DCR. The determination of the safety evaluation concluded that the DCR made changes to the facility as described in the UFSAR (Sections 3.8.3.1, 9.1.4.1, 9.1.4.2.c.4, 9.1.4.3.a 6, 9.1.4.7 b.2, and 9.1.4.4.1).

The safety evaluation concluded that the DCR was an enhancement and did not adversely affect the design basis or function of safety related structures, systems, or components.

The safety evaluation concluded that no unreviewed safety question was created by this DCR.

UFCR G7 019 19 of 32

- _ =.

MINOR MODIFICATIONS ~

The following Minor Modifications (MMOD) were installed during the period covered by this 10CFR50.59 Report. A safety evaluation was performed for each MMOD. Each - safety evaluation concluded that the MMOD did not create an unreviewed safety question.

Summaries of these MMODs and their~ safety evaluations are provided below.

MINOR MODIFICATION (MOD) NUMBER:

93 544 TITLE: Relief valves for WG C-5 A & B Bypass Lines

SUMMARY

DESCRIPTION AND PURPOSE:

This MMOD added relief valves to the PCCW side of the hydrogen gas compressors in the Waste Gas System. The addition of relief valves render the existing relief bypass lir.es obsolete. Therefore this moc'ification disconnected the relief bypass connections on the upstream side of the isolation valves. The existing check valves were also removed. The relief valves were added on the remaining portions of the rel.sf bypass lines thet are located on the downstream side of the isolations. This allows for maintenance of the hydrogen gas compressors during periods other than outages.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed on this MMOD. The determination of the safety evaluation concluded that the MMOD did make changes to the facility as described in the UFSAR (Figures 9.2-4 sh. 2 and 9.2-6 sh. 3). The safety evaluation concluded that the MMOD was an enhancement and did not adversely affect the design basis or function of any safety related systems, structures, or components. The safety evaluation concluded that no unroviewed safety question was created by this MMOD.

UFCR 94-042 MINOR MODIFICATION (MOD) NUMBER:

94-562 TITLE: SB Radiation Monitor

SUMMARY

DESCRIPTION AND PURPOSE:

.This MMOD rerouted the instrument tubing for the sample to the steam generator blowdown flash tank radiation monitor. This allows the sample to go through the cooler then the monitor. After leaving the monitor the sample goes through-the filter, then the pump and the flow switches. This modification allowed the monitor to see an unfiltered sample while the filters prevent magnetite from damaging the pump and fouling the flow switch.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed on this MMOD. The deterrnination of the safety evaluation applicability concluded that the MMOD made changes to the facility as 20 of 32

- MINOR MODIFICATION (MOD) NUMBER:

94-562 (Continued)

TITLE:-.SB Radiation Monitor

~ described in the' UFSAR (Section 11.5, Figure 10.4-11). The safety evaluation determined

~

that the MMOD was on enhancement and did not adversely affect the design basis or function of; safety related structures, systems, or components. The safety evaluation

)

concluded that no unreviewed cafety question was created by this MMOD.

UFCR 95-020 MISCELLANEOUS SAFETY EVALUATIONS:

The following miscellaneous safety evaluations were performed during the reporting period covered by this 10 CFR 50.59 Report. Each safety evaluation concluded that the situation or activity did not involve. an unreviewed safety question. Summaries of these safety evaluations are provided b; low.

MISCELLANEOUS UFSAR CHANGE:

UFCR 03-21 TITLE: N/A i

SUMMARY

DESCRIPTION AND PURPOSE:

This UFCR deleted UFSAR Table 9.5-9 " Listing of Unusually Hazardous Material," added an exception note to UFSAR Section 1.8, " Reg. Guide 1.70" and revised page F-98 of the BTP APCSB 9.5-1, Appendix A Report. The need for Table 9.5-9 no longer exists and the exception to Reg. Guide 1.70 describes the reason why the Table is no longer necessary.

The revision to page F-98 clarifies the three design requirements for the Fire Protection water storage tanks.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the changes made to the UFSAR by this UFCR, The need for 9ble 9.5-9 no longer exists. The table was required, prior to the promulgation of the current NRC tire protection requirements, to provide a listing of any unusually hazardous materials that could present unexpected fire hazards or otherwise complicate firefighting activities during a fire, 10CFR50, Appendix R requires

- that the identification of the type and location of all fire hazards and types of fires that could occur in each fire zone of the station be identified and that prefire strategies for fighting fires in safety related areas be developed. The combination of the requirements of Appendix R and the requirements of BTP APSCB 9.51, Appendix A are more detailed such that listing of hazardous materials separately on Table 9.51 is no longer necessary.

Separately, page F-98 of the Appendix A Report was revised to clarify the three design requirements of each of the Fire Protection water storage tanks. These requirements are:

tank capacity is 500,00 gallons e

300,000 gallons is reserved exclusively for fire protection and e

the minimum Technical Requirement water supply is 215,000 gallons.

21 of 32

m MISCELLANEOUS UFSAR C'HANGE:

UFCR 93 21'(Continued)

TITLE:1 N/A'.

The' safety evaluation concluded that these coanges do not involve changes._to any-

. equipment and there is no effect on the design basis or the operation of the Fire Protection System as a re:; ult'of these changes. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

MISCELLANEOUS UFSAR CHtNGE:

UFCR 95 37 TITLE: Extraction System Check Valve Closure Time / Testing i

SUMMARY

DESCRIPTION AND PURPOSE:

This UFCR deletes the two second closure time requirement for the extraction steam non-return check valves listed in UFSAR Section 10.2.2.3 and changes the frequency of testing

. for these valves as listed in Section 10.2.3.6 from daily to monthly. These check valvas close on a turbine trip and thus prevent stored energy in the feedwater heaters from flowing into the turbine and overspeeding it following the trip. The two second closure requirement and daily test are required of check valves in systems where " critical" potential energy exists in the heaters. No " critical" energy heaters exist so this UFCR deleted the two second closure requirement and modified the testing schedule.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support these. changes.

Critical" energy is defined by the turbine manufacturer (GE) as energy in a single heater greater than

- approximately one million BTUs, GE has confirmed that the hig*1est energy heater contains

-an energy of less than 100,00 BTUs, thus no critical heaters exist. The Extraction Steam

-System, of which the check valves are a part of, is non-safety and non-seismic. The safety evaluation concluded that the response of the plant followine a turbine trip with a failure of one or more of the check valves will not change the UFSAR accident ana!ysis associa'ed with turbine trip events. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

22 of 32 6

..mx

.-,w..

y

i MISCELLANEOUS UFSAR CHANGE:

UFCR 96 24 TITLE: Use of the Spent Fuel Bridge and Hoist

SUMMARY

DESCRIPilON AND PURPOSE:

Paragraph 9.1.4.2.c.2 of the UFMR states that the Spent Fuel Pool (SFP) bridge and hoist-are used " exclusively" for moving fuel assemblies in and around :he SFP area, in reality, it -

is necessary-to use the bridge and hoist to handle tools and components for core reconfiguration (RCCA, BPRA, Thimble Plug, Dummy Fuel Assembly, etc.). It is desirable to use the SFP bridge and hoist to handle irradiated debris containers and to support other fuel related maintenance activities'.

Lifts utilizing the SFP bridge and hoist are administratively controlled to ensure that the maximum load limit for the hoist is not exceeded. This UFCR revises text in UFSAR Section 9.1.4.2.c to clarify the use of th FP bridge and hoist for handling the additional items. Controls to ensure that an accidental drop of any of these loads will not exceed the bounds of the limiting case accident described in the UFSAR are also included.

GAFETY EW LUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the UFSAR changes made by this UFCR. The safety evaluation concluded that using the bFP bridge and hoist to lift and maneuver additional SFP tools and core components as well as using it to support fuel plated maintenance activities is acceptable. The basis for this conclusion is that all lifts using the SFP bridge and crane are administratively controlled for weight and lift height by normal station operating procedures. There is an overload protection device set to ensure that no load greater than the maximum load assumed in the UFSAR Chapter 15 accident analysis can be lifted. The loads described in this change will be lifted by trained personnel using qualified lifting components per the station approved Lifting Systems Manual. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

MISCELLANEOUS UFSAR CHANGE:

UFCR 96 28 TITLE: Resolution of UFSAR SFP Levels

SUMMARY

DESCRIPTION AND PURPOSE:

There are five significant water levels associated with the Spent Fuel Pool (SFH listed in the UFSAR. These water levels correspond to design basis levels that acep surface doses below regulatory and administrative dose limits during normal and abnormal situations in the SFP. The levels also describe the lowest physicallevel the SFP can reach as a result of inadvertent draining due to cooling and purification system pipe breaks. Minimum SFP level in order to comply with Technical Specifications is also listed.

There are two different sections in the UFSAR that discuss the various SFP levels. There were inconsistencies in these two sections for some of the SFP levels.

This UFCR reconciles these differences using approved calculations and controlled drawings to determine the correct SFP levels.

23 of 32

. MISCELLANEOUS UFSAR CHANGE:

UFCR 96-28 (Continued)

TITLE:. Resolution of UFSAR SFP Levels SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was comple.ted to support the UFSAR changes made to Sections 9.1 and 12.3 by this UFCR. The safety evaluation concluded that the changes made to the UFSAR did not affect the physical arrangement or procedures used to operate the plant. The different SFP levels were re-stated and validated. The dose rate from *.he minimum SFP levels for assumed accident conditions did not change.- No new equipment was added by these changes and no new methods of operating existing equipment were introduced. Technical Specification 3,3.11 provides the minimum SFP level when spent fuel is stored in the pool. This value _was unchanged. Technical Specification 5.6.2 is a plant features value for_ SFP level following a pipe break. This value will be revised independently of this UFCR. The safety evaluation concluded that no unreviewed safety

_ question was created by this UFCR.

MISCELLANEOUS UFSAR CHANGE:

UFCR 96-33 TITLE: Service Water Flow Rate UFSAR Update

SUMMARY

DESCRIPTION AND PURPOSE:

This UFCR updates UFSAR Tables 6.2-2, 9.2-1, 9.2-4, and 9.2-7 to be consistent with the latest calculation revisions that provide the besis for the data listed in those tables. The supporting calculations were revised under DCR 96 16, "PCCW Heat Exchanger Replacement." As a result of replacing the PCCW Heat Exchangers, the design basis values for Service Water flow, PCCW Not Positive Suction Head and PCCW flows were modified. The tables changed by this UFCR reflect the new values.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the UFSAR changes made by this UFCR. The changes to the data in the tables do not involve changes to plant design or operation. Replacement of the PCCW Heat Exchangers.had been previously evaluated in DCR 96-16 and the resultant changes to the design values had been found to be acceptable. This change incorporates the new design values. The revised design values demonstrate margin beyond system requirements. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

24 of.32

.~

i MISCELLANEOUS UFSAR CHANGE:

UFCR 9S 43 TITLE: ANSI Reg. Guide 3.1 Changes to UFSAR

(

SUMMARY

DESCRIPTION AND PURPOSE:

I The UFSAR referenced certain ANI/ ANSI guidelines when describing the qualification levels 9

of the operators enrolled in a licensed training program, if any deviation from the guidelines was rmcessary, a waiver request to the guidelines was submitted to the NRC on the license application. Seabrook Station's Licused Operator Training Program is accredited by INPO under the Systematic Approach to Training (SAT) process. Thus Seabrook Station no longer needs to request waive s fror, the NRC for license applicants who do not meet all of the ANI/ ANSI guidelines.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 scfety evaluation was completed to support the changes made to Sections 1.8,13.1, and Chapter 17, Appendix A of the UFSAR by this UFCR.

The changes constitute a change to procedures described in the UFSAR.

The safety evaluation described the lNPO accreditation and SAT processes and the NRC's acceptance of those processes as endorsed in the NRC's Final Policy Statement on Training and Qualification of Nuclear Power Plant Personnel. The Seabrook Station Licensed Operator Training Program is accredited by the National Nuclear Accrediting Board under this commitment. This accredita n qualifias Seabrook Station for the membership it maintains in the National Academy sei Nuclear Trcining. Acceeditation of training ensures that the Licensed Operator Training Program meets or exceeds tne requirements for a SAT as stated in 10 CFR 55, Sectbn 4. The programs qualify for the special status granted to systems based training by Section 31, Applications for License, and Section 50, Requirements, of 10CFR55. The safety evaluation concluded that by using ine SAT process North Atlantic will be able to determine the necessary qualifications for license applicants without having to request a waiver in Sect:on 12, Facility Operator Training Program, of the NRC license. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

MISCELLA'.'EOUS UFSAR CHANGE:

UFCR 96-66 TITLE: UFCR Section 9.1.5,

• Overhead Heavy Load Handling System"

SUMMARY

DESCRIPTION AND PURPOSE:

On April 11,1996, the NRC issued NRC Bulletin 96 02, " Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Cver Safe.y Related Equipment." North Atlantic responded to the Bulletin on May 10,1996, in the response, North Atlantic made a commitment to update the UFSAR to include "nn outline of tne details involved in the handling of heavy loads over spent fuel, fuel in the reacter core, or safety related equipment while the reactor was at power, (in all modes other than cold shutdown, refueling, er defueled). These changes are included in a new Section, 9.1.5 of the UFSAR and wer? mcorporated by this UFCR.

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MISCFLLANEOUS UFSAR CHANGE:

UFCR 96 66 (Continued)

TITLE: UFCR Section 9.1.5, *0 rerhead Heavy Load Handling System" SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the changes made to the UFSAR by this UFCR.

The additional information detailing the changes to the administrative controls that will be used when handling heavy loads in the situations described in NRC Bulletin 96 02, constitutes a change to procedures described in the UFSAR. The safety evaluation discussed the changes made to the administrative controls when handling NUREG-0612 type lifts that could be performed over spent fuel, fuelin the reactor core, or safety related equipment.

The safety evaluation referenced.the requirement that all NUREG 0012 type lifts that are to be made at power will have a 5(',59 safsty evaluation completed prior to the lift,if one has nnt previously been completed. The changes made by this UFCR make enhancements to North Atlantic's commitments to NUREG 0012 type lift requirements. The safety evaluation concluded that no unraviewed saiety question was created by this UFCR.

MISCELLANEOUS UFSAR CHANGE:

UFCR 97 02 TITLE: Editorial correction to UFSAR Section 10.4.8.6.c

SUMMARY

DESCRIPTION AND PURPOSE:

During the preparation of Minor Modification MMOD 95 639 it was identified that UFSAR Section 10.4.8.6.c "Demineralizer Subsystem" of the Steam Generator Blowdown System (SGBD) incorrectly states that the " pressure of the acid and caustic tanks are indicated."

P&lD 1 SB 20028 Rev.11 (UFSAR Figure 10.413 Sheet 2) shows that acid tank 1 SB TK-159 and caustic tank 1 SB TK 160 have level instruments but. do not have pressure instruments. Therefore, the acid and caustic tank pressurea are not indicated. This UFCR makes an editorial correction to UFSAR Section 10.4.8.6.c to remove the reference to pressure indication for the SGBD acid and caustic tanks. This makes the description of the instrumentation on the tanks as l' was originally installed.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support this UFCR.

The safety evaluation concluded that the change to the UFSAR was editorial in nature. The change does not alter the functionality or operation of the SGBD System. The use of the SGBD System during the UFSAR accidents in which it is discussed is not affected by this change.

The acid and caustic tanks are piped such that they cannot be pressurized so there is no need to have pressure indication. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

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MISCELLANEOUS UFSAR CHANGE:

UFCR 97 06 TITLE: Fire Protection Revisions i

SUMMARY

DESCRIPTION AND PURPOSE:

This UFCR makes changes to the UFSAR and Branch Technical Position (BTP) APCSB 9.51 Appendix A Report.

The changes are necessary to correctly account for pressure containing equipment that is in the control room and not listed in the UFSAR oection that discusses such equipment. Specifically not all of the fire extinguishers and self t ontained breathing apparatus and spare air cylinders that are in the control room were listed in the UFSAR.

This change also corrects the reference to Fire Protection Technical Specifications in the UFSAR.

The station was licensed with Fire Protection Technical Requirements, not Technical Specifications and this change corrects the UFSAR reference.

This UFCR a'so corrects various minor discrepancies in the Appendix A report such as the number of gas cylinders stored in safety related areas and the list of NFPA recommended i

equipment stored in each outside hose installation.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the changes implemented by this UFCR. UFSAR Sections 6.4.2.4b and 13.5.2.3.7a as well as Appendix A report pages F 62, 63 and 101 are edited. The safety evaluation concluded that this change will not degrade the ability of the Fire Protection System to perform its design function. The change doos not affect the operation, design basis or the failure modes and effects analysis of the Fire Protection System. The probability of an accident or malfunction of a different type than that which was previously evaluated in the UFSAR has not been created. The ability to achieve and maintain safe shutdown in the event of a fire has not been adversely affected by these changes. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

t 27 of 32

i l

MISCELLANEOUS UFSAR CHANGE:

UFCR 9714 TITLE: Exceptions to Single Failure Criteria During Channel Bypass for Maintenance and Testing

SUMMARY

DESCRIPTION AND PURPOSE:

This UFCR changed UFSAR discussions relating to codes and standard:. associated with exception to the single failure criteria in IEEE Standard 270 1971, Paragraph 4.11. The exception is applicable to certain Solid State Protective System (SSPS) funct5ns that are two out of three and two out of four logic during channel bypass for maintenance or testing.

L DCR 94 02 installed test circuits that permitted the bypassing of additional channels for maintenanca or testing. Bypassing certain SSPS channels does not leave the sufficient number of channels to meet.the IEEE 279 1971 single failure criteria which the original plant Safety Evaluation Report (SER) and FSAR assumed. However, SERs completed by the NRC in their review of the channel bypass design change, and the safety evaluations submitted to support the associated license change, accepted channel bypass for maintenance and testing.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the UFSAR changes implemented by this UFCR. The safety evaluation concluded that the NRC accepted the conclusions of WCAP 10271 in approving the use of channel bypass during maintenance and testing. WCAP-10271 demonstrated acceptable protection system reliability for the 2/3 and 2/4 channels during channel bypass for maintenance and testing. One stipulation required for approval was that the use of channel bypass did not rely on the lifting of leads or the installation of jumpers in order tu achieve the channel bypass. The design change implemented in 1995 met this condition.

The safety evaluation performed to support the UFSAR changes to Sections 3.1, 7.1 and 7.2, implemented by this UFCR, further stated that the proposed changes to the single failure criteria do not involve a change to the function or failure modes of any part of the facility. The structural and functional integrity of tne reactor protection and engineered safety features actuation systems is unaffected..The safety evaluations presented to and approved by the NRC when the bypass testing desigt change was implemented, provided the justification and basis for the change. The net eff sct of the change was determined to be an increase in reactor safety. The changes madi by this UFCR correct an oversight made during the implementation of the channel bypars design change. The wording in the UFSAR is changed to explicitly take exception to th; single failure criterion for 2/3 and 2/4 SSPS channels during use of channel bypass durint maintenance and testing. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

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i MISCELLANEOUS UFSAR CHANUE:

UFCR 9718 TITLE: None i

SUMMARY

DESCRIPTION AND PURPOSE:

This UFCR clarified unintentional restrictive wording in UFSAR Section 11.4.2.3.

That section vescribes the Resin Slulce Tanks and states that the tanks have a low pressure nitrogen overpressure on them "at all times." It is not necessary to maintain a nitrogen overpressure on the tanks during resin changeout or maintenance on the tanks and tank components. During these times the tanks must be vented, purged or ventilated in order for the resin replacement or maintenance to be conducted. The changes enacted by th;s UFCR clarified when a nitregen overpressure was necessary on the Resin Sluice Tanks.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the changes to the UFSAR.

UFSAR Sectior.11.4.2.3 was changed to remove the "at all times" requirement of a nitrogen overpressure on the tanks. The safety evaluation concluded that the nitrogen alignment to the Resin Sluice Tanks is not an initiating factor in any of the radoactive releases from a system or component previously identified in any accident described in the UFSAR. Accidents evaluated in the UFSAR are the rupture of one Resin Sluice Tank, a Floor Drain Tank or a Boron Waste Storage Tank. The postulated rupture of the Resin Sluice Tank is not affected by the presence or lack of a nitrogen overpressure. The nitrogen piping and the Resin Sluice Tanks are in the Waste Processing Building and are not nelied on to support the operation of any systems important to safety, No new failure mechanisms are created by operating the Resin Sluice Tanks in the manner described in this change. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

MISCELLANEOUS UFSAR CHANGE:

UFCR 97 26 TITLE: Operation of the Letdown Degasifier to the Aerated Vent Header

SUMMARY

DESCRIPTION AND PURPOSE:

Operating the Letdown Degasifier while it is aligned to the aerated vent header was 1

necessary during the pre refueling period following the plant shutdown in preparation for refueling outage OR05. The Letdown Degasifier is normally operated when aligned to the aerated vent header during the post refueling startup phase of plant operation.

The existence of increased concentrations of radioactive gasses in the Reactor Coolant System (RCS), caused by fuel failures that occurred during the operating cycle required as much degassification as possible, for as long as possible, prior to opening the RCS for refueling and maintenance.

Extended operation of the Letdown Degasifier in this way was necessary to limit the exposure of plant workers to the radioactive gasses in the RCS from the failed fuel. This reduction of exposure was in keeping with the ALARA goals of the station.

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1 MISCELLANEOUS UFSAR CHANGE:

UFCR 97 26 (Continued)

TULO Operation of the Letdown Degavfier to the Aerbted Vent Header l

C/geon of the Letdown Degasifier while lined up to the aerated vent header was necessary because of the potential for admitting oxygen into the Hydrogenated Waste Gas l

l System. That might have created an explosive mixture of hydrogen and oxygen. This necessitated lining up the Letdown Degasifier discharge to the aerated vent header and i

sending it through the plant vent HEPA filter prior to being discharged out the plant vent.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support this UFCR. UFSAR Section 9.3.4.2.e.11 discusses running the Letdown Degasifier while lined up to the aerated vent header during a post refueling condition. UFSAR Section 12.2.4.4 states that noble gasses are removed frorn the letdown purification flow by the Letdown Degasifier and are processed by the Radioactive Waste Gas System. Both of these UFSAR se:tions were changed as a result of the, change to the operation of the Letdown Degasifier.

The safety evaluation concluded that operation of the Letdown Degasifier in the specified manner would have no adverse impact on the Letdown System or the Waste Gas System.

The gesses discharged through the aerated vent header to the plant vent HEPA filter and ultimately out the plant vent were monitored by the Wide Range Gas Monitor (WRGM) to ensure compliance with the Technical Specification limits for offsite releases. Sampling of the Letdown Degasifier effluent prior to discharging it to the aerated vent header ensured that the concentretion of hydrogen in the effluent was below an explosive concentration.

The safety evaluation concluded that there was no impact on any system required to support safe shutdown of the plant as a result of the changes. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

MISCELLANEOUS UFSAR CHANGE:

UFCR 97 20 TITLE: Fuel Reconstitution

SUMMARY

DESCRIPTION AND PURPOSE:

Seabrook Station experienced fuel rod degradation during the last operating cycle. The degraded fuel rods were repaireo by using a process called fuel reconstitution.

Fuel reconstitution is accomplished by partially disassembling the fuel rod and replacing the leaking fuel pins with stainless steel filler rods. The reconstituted fuel rods are expected to retain the same functional and structural characteristics as the pre-worked assemblies or assemblies with equal operating history. This UFCR made the UFSAR changes necessary to permit the use of ~ the special procedure developed to conduct fuel reconstitution at

Seabrook Station and demonstrated that the use of this procedure did not subject the plant to an unsafe condition.

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MISCELLANEOUS UFSAR CHANGE:

UFCR 97 29 (Continued)

TITLE: Fuel Reconstitution SAFETY EVALUAYlON

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the UFSAR changes made to Section 9.1.4.2.c.5 by this UFCR Changes to the UFSAR in order to allow the use of the new fuel elevator during fuel reconstitution were necessary.

The safety evaluation referenced NRC Generic Letter (GL) 90-02 that proposed recommended changes to plant Technical Specifications to permit fuel assembly reconstitution. Seabrook Station fuel was designed to be able to be reconstituted should the need arise. Procedure FP-NAH FR1 "Seabrook Fuel Reconstitution [RTN)" provides a detailed, safe and effective method to

}

conduct Removable Top Nozzle (RTN) fuel reconstitution operations.

The procedure includes designation of responsibi'itles, precautions and prerequisites for fuel reconstitution.

lt also provides instructions to execute all phases of the fuel reconstitution.

The :afety evaluation concluded that the consequence of dropping any tool used during the fuel reconstitution was bounded by the existing UFSAR analysis for a dropped fuel assembly. The design basis fuel handling accident offsite dose analysis is based on the Reg. Guide 1.25 assumption of a water depth of 23 feet above the damaged fuel assembly. Since the reconstitution took place at a depth of 10 feet of water above the fuel assembly, the fuel handling accident was re-evaluated with the assumption that damage occurs to the assembly at a water depth of 10 feet during fuel reconstitution.

The only fuel handling-accident analysis parameter that was affected by this change in water depth was the iodine pool scrubbing factor. The original UFSAR analysis uses a scrubbing factor of 100. This is conservatively less than the actual scrubbing factor for 23 feet of water. The calculated scrubbing factor for 10 feet of water is 122 so the actual scrubbing factor for 10 feet is still conservative with respect to the original analysis and thus the original analysis is still bounding.

The safety analysis concluded that because the fuel reconstitution procedure did not affect the function of equipment or systems used for accident mitigation, because no operating conditions of safety related equipment were altered. The safety evaluation concluded that no unroviewed safety question was created by this UFCR.

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~

l 1

MISCELLANEOUS UFSAR CHANGE:

UFCR 97 31 TITLE: Reactor Coolant Lithium Control l

SUMMARY

DESCRIPTION AND PURPOSE:

Tiie Reactor Coolant System (RCS) lithium control program was changed in order to raise Beginning Of Life (BOL) RCS pH thereby reducing corrosion in the RCS. Reduced corrosion is expected to create an RCS environment that is less favorable to corrosion product (CRUD) deposition on the fuel cladding surfaces. The change is supported by EPRI studies that have recommended the change for extended cycle, high temperature plants like Seabrook Station that have experienced axial offset anomaly.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the change in the RCS lithium control program. The RCS chemistry program is described in UFSAR Section 5.2.3.2.a.

UFSAR Table 5.2 5 references chemistry specifications and limits. The change to the lithium control program increased the lithium concentration above the limit specified in l

Table 5.2 5. The safety evaluation concluded that based on a Seabrook Station specific evaluation, the integrity of the materials contained in the RCS are not challenged by the change in RCS lithium concentration or the associated change in RCS pH.

The evaluation specifically evaluated the effects the RCS chemistry changes would have on RCS bounday surfaces and components as well as the systems and components which interface with the RCS including ECCS systems and components. The evaluation conclude that any system or component material effects would be within design specifications. The safety evaluation concluded that no unreviewed safety question was created by this UFCR.

.t L

32 of 32

LNCLOSURill_I{tMLN.97125 Summaries of the Remaining 10CI'R50.59 Safety Evaluations Performed for Temporary Modifications, Procedure Changes.

Commitment Changes, and Miscellaneous Changes

DESIGN CHANGE RECORD (DCR) NUMBER:

89 062 TITLE: Update Switchyard Alarms and Settings

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR resolves a number of minor 345 KV switchyard / relay room issues resulting in

-improved control, protection, and alem features associated with the offsite power supply.

The changes include time delay of nuisance alarms, deletion of unrequired and nonfunctioning alarms, correction of wiring connections to attain an intended alarm function, removal of controls and alarms for unused Unit 2 components and a setting refinement for gas circuit breakers D and G reclosing relays.

SAFETY EVALUATION

SUMMARY

' A 10CFR50.59 evaluation was performed for this DCR. The determination of safety evaluation applicability concluded that the DCR made changes to the f acility as described in the UFSAR. The DCR affected Chapter 8. Design and safety evaluations dependent upon these provisions remain valid, no acceptance levels or systern limitations have been altered and no design or safety margins have been reduced by these changes. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

DESIGN CHANGE RECORD (DCR) NUMBER:

92 048 1

TITLE: Piano Hinge Replacement

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR was issued to replace the plano hinges on fourteen "special" doors. These doors are "special" due to their functional requirements, which include bullet resistant, seismic category 1, tornado pressure resistance, three hour fire rating, etc. The hinges installed on these doors are a continuous " piano" type. This hinge design has been experiencing significant wear due to the high usage of these doors.

The proposed alternative design is a leaf type hinge that will support the door weight, meet functional requirements as well as be able to stand up to high usage. The design, function and service conditions that the leaf hinges are qualified for are identical to that of the existing hinges.

SAFETY EVALUATION

SUMMARY

A.10CFR50.59 safety evaluation was performed for this DCR The safety evaluation applicability concluded that the DCR did make changes to the facility as described in the UFSAR (Sections 3.3.2 and 6.5.1). This modification is a plant maintenance and operational enhancement. Design and safety evaluations dependent upon these provisions remain valid, no acceptance levels or system limitations have been altered and no design or safety margins have been reduced. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

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1 DESIGN CHANGE RECORD (DCR) NUMBER:

94-022 TITLE: North Road Access Control System

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR provided a North Access Road Control System to reduce the need for the Security Department to staff the North Access Road Gatehouse during off peak hours, weekends and holidays.

Access is provided by an automated card entry system, implementation of this DCR involved equipment installation along and near the North Access Road and inside the Protected Area Gatehouse.

i SAFETY EVALUATION

SUMMARY

' A 10CFR 50.59 evaluation was performed for this DCR, The determination of safety evaluation applicability concluded that the DCR did make changes to the facility as described in the UFSAR. The design change does not alter the physical plant facilities as described in the UFSAR, but did require a review of the Emergency Diesel Generator loading to ensure that the additional loads were within rating of the diesel. Distribution panel loading and circuit protection coordination was reviewed and found acceptable. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

DESIGN CHANGE RECORD (DCR) NUMBER:

94 037 TITLE: Erosion / Corrosion Program Large Bore Piping Component Replacement

SUMMARY

DESCRIPTION AND PURPOSE:

Thir DCR replaced existing non ASME, non safety related carbon steel piping and components that have been projected to have unacceptable wall thickness due to erosion or corrosion. The originalissue replaces piping which may not meet or exceed the required minirnum wall thickness at the fourth refueling outage, thus maintaining structural integrity of each system during the next cycle.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this DCR. The determ! nation of the safety evaluation applicability concluded that the DCR made changes to the facility as described in the UFSAR (Sections 3.9,10.3, and 10.4). Design and safety evaluations dependent on these provisions remain valid, no acceptance levels or system limitations have been altered and no design or safety margins have been reduced. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

2 of 46

DESIGN CHANGE RECORD (DCR) NUMBER:

95 005 TITLE: Oil / Water Separator Vault (OWSV) 1&2 LevelInstrument Replacement

SUMMARY

DESCRIPTIO"i AND PURPOSE:

This DCR replaced C.s two level switches that contrcl the sump pumps and alarm outputs with a single ultrasonic level transmitter and pump controller in both OWSV #1 and #2. The i

alarm setpoints were also revised.

l SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this DCR. The determination of the safety evaluation applicability concluded that the DCR made changes to the facility as described in the UFSAR (Sections 7, 8, 9 and 9.3.3). Design and safety evaluations dependent upon these provisions remain valid, no acceptance levels or system limitations have been altered and no design or safety margins have been reduced. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

DESIGN CHANGE RECORD (DCR) NUMBER:

96 022 TITLE:

Replacement of Waste Distillate Flowmoter 1 WL-FE/ FIT 14581

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR replaced the existing Fischer Porter Mag Flowmoters l WL-FE 14581 and 1 WL-FIT-14581 by an inline Foxboro vortex shedding flowmeter model 83F. The new flowmeter is an inline flow element with an integral flow transmitter / local indicating module, and has the same functional capability of the flowmeters it replaced. It was tagged I WL FIT 1458-1.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was prepared to support this DCR. The WL System instrumentation is functionally described in Section 11.2.6. of the UFSAR. The description is general in nature; no individual control loop is described. This DCR does not change the functiona! capability of control loop 1458. The final control elements FCV-14581 will still operate as designed i.e., fail safe in the closed position. It is only a change in usage of a different technology. Thus, UFSAR text is unaffected by this DCR. However, since the replacement inline Foxboro vortex shedding meter provides the combined function of the flowmeters 1 WL FE 14581 and 1 WL FIT 14581, P&lD WL 20831 and UFSAR Figure 11-24 were affected.

Accident analysis described in UFSAR Chapter 15.7.2 is unaffected by this DCR. The improved accuracy of the new flowmeter will not compromise the monitoring of flow of 3 of 46 I

l

-.=.

= -

DESIGN CHANGE RECORD (DCR) NUMBER:

06-022 (Continued)

TITLE:

Replacement of Waste Distillate FlowrMer 1 WL FE/ FIT 14581 effluent releases to the environment. The installatlon is Non Seismic and NNS Safety Class.

Malfunctions of equipment important to safety such as radiation element RE6509, FCV.

14581, FCV 1458 2, etc. due to seismk. ally generated missiles will not be any more probable as a result of the DCR changes than the previous installation. The failure mooo of the valve is unaffected by this DCR. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

DESIGN CHANGE RECORD (DCR) NUMBER:

97 017 TITLE: Main Steam Line Radiation Monitor Relocation

SUMMARY

DESCRIPTION AND PURPOSE:

The Main Steam Line Radiation Monitor Detectors were originally installed in a location that prevented them from monitoring the steam flow through the Atmospheric Steam Dump t

Valves (ASDV). Since this results in an unmonitored release, that location for the detectors is unacceptable. Steam released through the ASDVs was not monitored for detectable radiation by the monitors in the main steam line downstream of the ASDVs. This DCR relocated the monitors and placed them approximately 2 feet upstream of the ASDVs. At that location they will be able to monitor steam released through the ADVs as well as the steam flowing to the turbine generator and steam supplied auxiliary systems.

SAFETY EVALUAT!ON

SUMMARY

A 50.59 safety evaluation was completed to support this DCR. UFSAR Section 11.5.2.1.]

states that the main steam line radiation monitors channel detectors shall be located upstream of the safety relief salves. The original detector installation location did not comply with that requirement. Since this DCR relocated the monitors discussed in the UFSAR it was considered a change to the facility as described in the UFSAR. The monitors continue to meet the required environmental, design and seismic regulations imposed on post accident monitoring radiation detectors. The sof tware used to calculate on and offsite dose projections has been revised to accommodate the new detector locations. The safety evaluation concluded that no unreviewed safety question was created by this DCR.

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MINOR MODIFICATION (MOD) NUMBER:

90 630 TITLE: Vibration Reduction of Miscellaneous Elements Associated with MSR Control Panel MS-CP 73 & Heaters CO E 25A/B & FW E 26A/B (CA2 7,13,1%

SUMMARY

DESCRIPTION AND PURPOSE:

This MMOD provided revisions, modifications, and replacements to supports for piping in the MS, EHC, CO, FW, and miscellaneous secondary systems. These changes are being implemented to reduce vibration induced system stresses.

SAFETY EVALUATION

SUMMARY

A'10CFR50.59 evaluation was performed for this MMOD The determination of the safety evaluation applicability concluded that the MMOD made changes in the f acility as described in the UFSAR. The safety evaluation concluded that the MMOD was an enhancement and did not adversely affect the design basis or function of safety related structures, systems, or components. The safety evaluation concluded that no unreviewed safety question was created by this MMOD.

MINOR MODIFICATION (MOD) NUMBER:

91 513 TITLE: Fuse Changes and Lamp Circuit Board Modifications for Inverter 1 EDE l 1E and 1 EDE l 1F

SUMMARY

DESCRIPTION AND PURPOSE:

This MMOD implemented minor changes to two circuit boards within inverters 1 EDE l 1E and 1 EDE t 1F. These changes were required to improve the reliability of the precharge circuit and to provide lamp indication upon initiation of the precharge push-button. This MMOD also corrected vendor documentation errors.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed on this MMOD. The determination of the safety evaluation applicability concluded that the MMOD made changes in the f acility as described in the UFSAR, Chapters 1 and 8. These modifications correct a minor hardware error in order to upgrado equipment to original expected performance and had no affect on current associated analysis. The safety evaluation concluded that no unreviewed safety question was created by this MMOD.

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_ _ - ~

P MINOR MODIFICATION (MOD) NUMBER:

91 632 TITLE: Spent Fuel Bridge and Holst (1 FH RE 2) Load Test Fixture Location

SUMMARY

DESCRIPTION AND PURPOSE:

This MMOD relocated the load test fixture to an area where the crane wheels and raila will make proper contact. When calibrating the load cell on the fuel handling crano for crane interlock testing, inaccurate data is obtained due to one wheel not fully contacting the rail when parked at the load test fixture location.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this MMOD. The determination of the safety evaluation applicability concluded that the MMOD made changes to the facility as described in the UFSAR (Sections 1.2.2.5, 3.8.4.1(h), 9.1, 9.1.2.2, and Figure 1.215 through 21), The load test fixture satisfies all structural engineering acceptance criteria.

The safety evaluation concluded that the MMOD did not adversely affect the design basis or function and was included in the current analyses. The safety evaluation concluded that no unreviewed safety question was created by this MMOD.

MINOR MODIFICATION (MOD) NUMBER:

93 548

' TITLE: Alterex B ush Rigging Modification

SUMMARY

DESCRIPTION AND PURPOSE:

This MMOD changed the nurnber of Attorex Exciter brushes (reduced the number to 2),

adjusted the tension on the brush rigging, and installed new brush rigging that incorporates enhancements for improved performance.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 was performed for this MMOD. The determination of the safety evaluation applicability concluded that the MMOD made changes il the facility as described in the UFSAR. The main generator is described in the UFSAR Sections 1.2.9.2, 8.2.1.3.2.b, and 10.2.2. The safety evaluation concluded that the MMOD did not udversely affect the design basis or function and was included in the current analyses. The safety evaluation concluded that no unreviewed safety question was created by this MMDD.

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MINOR MODIFICATION (MMOD) NUMBER:

91 570 TITLE: Containment Building Polar Crane Safe Load Path Drawing Revision

SUMMARY

DESCRIPTION AND PURPOSE:

To allow for better utilization of floor space dunng refueling outages, Drawing 805054 (Containment Structure Refueling Equipment Laydown General Arrangement) was revised to accommodate the location of temporarily relocated equipment and associated services.

Due to the revised laydown drawing and load handling operatiens associated with outages, fou' (4) safe load path drawings for control of heavy loads carried by the Containment Polar Crane, required revision.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was prepared to support this MMOD. Safe load path drawings are one of a number of controls placed on the movement of heavy loads in the containment.

Expansion of the safe load paths allowed by these drawings maintain applicable design criteria and qualifications for control of heavy loads in the containment.

The basis for the original qualification of the Polar Crane is unaffected as a result of safe load path drawing revisions issued via this MMOD. The safety evaluation concluded that no unreviewed safety question was created by this MMOD.

MINOR MODIFICATION (MOD) NUMBER:

91 601 TITLE: Replacement of 1/P Converters for Pressurizer Spray and Boric Acid Blend Valves

SUMMARY

DESCRIPTION AND PURPOSE:

This modification replaced Fisher current to pneumatic l/P transducers with Rosemount 1/P transducers in the control loops for the Pressurizer Spray Valves. In addition, the existing valve positioncrs for the Boric Acid Addition Inlet Valve and the total RMW Addition Flow Valves wers replaced by a combination 1/P transducer with built in positioners manufactured by Siemens. Tl.e Fisher I/P transducers were replaced since they continually drif ted out of tolerance.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this modification. The determination of safety evaluation applicability concluded that the MMOD made changes in the facility as described in the UFSAR. The components differ slightly in fit and form, but there is no adverse effect on the system's ability to perform its function and no new failure modes were created. The alternate transducers and positioner models will equal or exceed the critical performance characteristics of the original equipment. The Pressurizer Pressure Control. System is discussed in.UFSAR Section 7.7 and the Boric Acid Makeup Control

System is discussed in UFSt.R Section 9.3.

The safety evaluation concluded that no unreviewed safety question was created by this MMOD.

7 of 46

TECHNICAL REQUIREMENT (TR) CHANGE: -

TR NUMBER 96 03 TITLE: Loose Part Monitoring System

SUMMARY

DESCRIPTION AND PURPOSE:

TR 96-03 revised Technical Requirement (TR) 3,

• Loose Part Detection System" to permit the use of normally passive Loose Part Monnoring System (LPMS) sensors on each Steam i

Generator (SG) to meet operability requirements. The passive sensor on each SG is located in close ploximity to the two active sensors and while it is normally used as an additional aid in loose part localization, it meets the same Reg. Guide 1.133 requirements as the two normally active sensors and thus can be subctituted for one of the active sensors should one of the active sensors fail. TR 96 03 also clarifies that the Channel Check and Analog Channel Operational Test surveillance requirements only pertain to the active channels or to a passive channelif it was being substituted for an active channel.

SAFETY EVALUATION

SUMMARY

A 10 CFR 50.59 cvaluation was performed for this change.

The safety evaluation concluded that the normally passive LPMS sensor is capable of providing the same function as one of the active sensors on the SG because the field hardware and mounting configuration are identical; when switched into active service the passive sensor utillres cabinet electronics of one of the normally active sensors on another SG; alarm setpoints for all LPMS cabinet modules are set to the same value; the passive sensor monitors the SG reactor coolant intet plenum similar to the active sensors; and the passive sensors meet all the same design requirements detailed in Reg. Guide 1.133 " System Characteristics" as the active sensors do. The safety evaluation concluded that no unreviewed safety question was created by this TR change.

TECHNICAL REQUIREMENT (TR) CHANGE:

TR NUMBER 96 04 TITLE: Safety Evaluation for incorporation of Incore Detector, Seismic, Meteorological and Turbine Overspeed Protection Instrumentation-Related Technical Requirements into the SSTR.

SUMMARY

DESCRIPTION AND PURPOSE:

NRC Generic Letter 9510 " Relocation of Selected Technical Specifications Requirements Related to Instrumentation" and the NRC " Final Policy Statement on Technical Specifications improvements for Nuclear Power Reactors" (58 CFR 39132) allowed the relocation of certain Technical Specifications into the Technical Requirements Manual. The

. Technical Specifications relocated to the Technical Requirements Manual by this change are not related to dominant contributors to plant risk and thus do not meet the criteria established by the NRC (10 CFR 50.30) for determining design conditions and associated surveillances that should be included in Technical Specifications. Savings of both costs and time will be reallied by both the NRC and North Atlantic as a result of moving these requirements to the licensee controlled program.

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1 TECHNICAL RL MEMENT (TR) CHANGE:

TR NUMBER 96 04 (Contint,ed)

TITLE: Safety Evaluation for incorporation of incore Detector, Seismic, Meteorological and Turbine Overspeed Protection Instrumentation Related Technical Requirements into the SSTR.

SAFETY EVALUATION

SUMMARY

A 10 CFR 50.59' evaluation was performed for this change.

The safety evaluation concluded that the systems listed in this change are not part of an initial _ condition of a design basis accident or transient that either assumes the failure (,f or presents a challenge to the integrity of a fission product barrier. They are not relied upon as a primary success path to mitigate design basis accidents or transients. The proposed changes do not involve any physical changes to the plant, do not alter the way eny structure, system or component functions, do not modify the manner in which the plant is operated and do not impact the physical protective boundaries of the plant.

The instrumentation-related systems listed in this change will continue tc. be administratively controlled in accordance with Technical Specification Section 6.0

" Administrative Controls." The changes do not degrade the ability of systems, structures or components important to safety of performing their safety functions. The change will not change the response of any system, structure, or component important to safety as described in the UFSAR.

The safety evaluation concluded that no unreviewed safety question was created by this TR change.

1 9 of 46 I

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STAND ALONE 50.59 SAFETY EVALUATIONS:

I i

EVALUATION NUMBERS: ACR 97 0094; ACR 97 0181; ACR-97 0210 TITLES:

97 0094 Partially Opening Valve SI V228 to Eliminate SIS Test Line Overpressure Condition 97-0181 Openir.,

Valve SF V 100 to Eliminate Refueling Canal Transfer Line Overpressurization 97 0210-Opening Valve WLD V79 to Eliminato RCDT to PDT Overpressure Condition for i

WLD V 81

SUMMARY

DESCRIPTION AND PURPOSE:

- During the reviews fos NRC Genetic Letter (GL) 96 06, Assurance of Equipment Operability and Containment integrity During Design Basis Accident Conditions, several inside containment isolation valves could be challenged due to thermally induced overpressure conditions in piping that is connected to the containtnent isolation valves.

The linos identified as being susceptc.~.s were the Safety injection System leakoff test line;

}

the refueling canal drain line; and the Reactor Coolant Drain Tank (RCDT) to Primary Drain Tank (PDT) transfer line. Each of these conditions were part of the original design of the systems at the time of the initiallicensing of the plant.

One hour non emergency reports were made to the NRC informing them of the conditions.

1 Appropriate Technical Specification action statements were entered and short term corrective actions with supporting operability determinations restored the affected equinment to operable status. Long term corrective actions included installation of relief valves and modifying certain valvo line ups to provide overpressure protection for the affected lines. Separate 50.59 evaluations were completed for the long term corrective actions.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was completed to evaluate short term corrective actions put in place to restore operability to the affected lines and associated containment isolation valves.

The opening of valves in order to provide a relief path that will prevent overpressur9ation of the affected lines durhg post LOCA conditions was ovaluated and found to be an effective short term corrective action for two of the three lir'es. The safety evaluation concluded that opening a drain valve and allovsing small amounts of water to drain to the containment ' drain sump did not create an increase in the probability or consequences of any accident previously evaluated or the malfunction of any equipment important to safety.' Realign!ng the line-up of the RCDT to PDT transfer line and manually pumping down the RCDT when necessary provided the overpressure protection for that

- line.

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STAND ALONE 50.59 SAFETY EVALUATIONS:

(Continued)

EVALUATION NUMBERS: ACR 97 0094; ACR 97 0181; ACR 97 0210 Mir imizing the use of the affected lines and stationing a dedicated operator to provide

[

overpressure protection for the RCDT line should a LOCA occur during the RCDT to PDT f

transfer s'so helped mitigate the overpressurization concern and the minor effects of the corrective actions. The three safety rw 1ations concluded that no unreviewed safety questions were created by these change...

STAND ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER 97 048 001 r

- TITL2: Dry Run of Fuol Transfer Cart in Support of 96 WOO 2953

SUMMARY

DESCRIPTION AND PURPOSE:

The Socrn ild mlinal handling transfer drive system and controls were replaced during refueling outm9 SrfGB in May of 1997. In order to test the newly installed equipment a

" dry run" of thi, equipment was necessary. The "dr*/ ori" consisted of closing the spent fuel pool gate and dewate:Ing the cask handling area and the fuel transfer area in the Spent Fuel Building. The planned testing then included operating the fuel transfer cart between the Fuel Storage Building and the Containment Building to test and adjust as necessary the limit switches associated with upender oposation.

N'>rt'1 Atlantic's response to INPO SOER 851 dascribing loss of eactor cavity or spent fuel poolinventory includad the implementation of severa! recommendations that were aimed at preventing and mh.9ating such events at Seabrook Station.

One of the implemented recommendations would prevent the " dry run" test. The recommendation administratively prohibited the opening of the fuel transfer valve unless the spent fuel transfer system and reactor cavity water levels were at the same height as the spent fuel pool level. The purpose of Safety Evaluation 97 048-001 was to take a one time exception to the plohibition of opening the fuel transfer valve without the water levels in the two buildings at equal height and to provide the administrative measures necessary to enable the " dry run" to be coriducted safely.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed for proposed test. The safety evaluation determined that the UFSAR recognized that maintenance would have to be conducted periodically on the refueling equipment and that tho fu6l handling canal would have to be dowatered to perform such maintenance. The UFSAR also described a pre-operational " dry

-tun" of the fuel transfer system required for the initial check out of the original equipment.

The - UFSAR does not contain language that reveals the understanding of a possible replacement of the fuel transfer equipment and thus a future " dry run" necessary to test such equ:pment. For this reason the safety evaluation concluded that the " dry run" of the new fuel transfer uquipment was a test not described in the UFSAR.

.11 of 46

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I

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i STAND ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER 97 048-001 (Continued)

TITLE: Dry Run of Fuel Transfer Cart in Support of 96 WOO 2953 The safety evaluation reviewed the Chapter 15 accident analyses in the UFSAR and determined that the " dry run" test would not affect the fuel handling accident in the Spent Fuel Building or the Spent Fuel Cask Drop accident as described in 'he UFSAR because the test would be conducted in Mode 5 with the reactor vessel head in place and spent fuel movement o9erations suspended during the test, f

A separate 10CFR50.59 safety evaluation evalue'ed the replacement fuel transfer i

equipment and determined that no unreviewed safety question existed as a result of the change in equipment. The safety evaluation for the " dry run" concluded that no changes were being made to any other plant equipment as a result of the " dry run test. Therefore the probability or consequences of a malfunction of equipment important to safety were not increased as a result of the " dry run" test.

The spent fuel pool gate seats were addressed in detail in the safety evaluation, it was concluded that the catastrophic failure of the seals was not considered likely because the seats were new having been replaced in July of 1996; there are check valves in the air supply lines to the seals designed to maintain seal pressure even on a loss of air; the air check valves were tested prior to dowatering; the design of the seal housing arrangement has the seal enclosed in four stainless steel walls which contribute to seal integrity; the cask handling area will have been dewatered for four weeks prior to the " dry run" test and the integrity of the seats will have been monitored prior to the test. The safety evaluation concluded that no enreviewed safety question was created by this test.

STAND ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER:

97 041 001 TITLE: Modified EOL Moderator Temperature Coefficient

SUMMARY

DESCRIPTION AND PURPOSE:

This 10CFR50.59 safety evaluation was written to support a change to the Cycle 5 Core Operating Limits Report (COLR). The change to the COLR was necessary because of the Cycle 5 Moderator Tcmperature CoeffMent (MTC) measurements performed, as required by the Technical Specifications, after reaching an equilibrium boron concentration of 300

' PPM. These MTC measurements indicated that the 300 PPM surveillance limit was not met.- Technical Specifications require MTC to be remeasured at least once per 14 Effective Full Power Days (EFPD) when the measured MTC is more negative than the 300 PPM limit. Since the MTC measurement is a difficult surveillance to conduct, a change to the value of the 300 PPM MTC limit contained in the COLR was performed.

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1 STAND ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER:

97 041 001 (Continuedl TITLE: Modified EOL Moderator Temperature Coefficient l

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support the change in the 300 PPM MTC COLR limit'. The safety evaluation concluded that the change was justified based on the Cycle 5 safety analysis. Two analyses required revisions to specifically evaluate the impact of the change in the 300 PPM MTC value. - These analyses were the Main Steam Line Break Analysis and the Verifictition of Borated Water Source Requirements.. The revisions to these analyses provided justification for operation of the remainder of Cycle 5 with a modified 300 PPM MTC value.

The 10CFR50.59 safety evaluation considered the effect that the more negative than expected MTC value had on shutdown margin boron results.

Previously determined shutdown margin data are verified axplicitly during startup physics testing and by continuously comparing the boron lotdown curve to predicted boron concentration during cycle operation. All comparir,ons of boron concentration predictions to measured data are within the expected uncertainty. Thus, the safety evaluation concluded, there is no impact of the observed discrepancy between the predicted and measured MYC values on the remaining safety analyses parameters.

The 10CFR50.59 safety evaluation concluded that the proposed changes to the Cycle 5 COLR will allow core operation with a modified 300 PPM All Rods Out (ARO), Rated Thermal Power conditions MTC value while ensuring plant safety. This conclusion is based on the existing and re analyzed Cycle 5 safety analyses completed to support the change.

The safety evaluation concluded that no unreviewed safety question was created by this change.

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STAND ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER: 96-061 001 TITLE: Use of SFP Bridge and Holst to Handle the SFP Gate Seal Assembly

SUMMARY

DESCRIPTION AND PURPOSE:

The Spent Fuel Pool (SFP) gate has an inflatable seal that had reached the end of service life and needed replacement. The UFSAR described the SFP Bridge and Holst in Section 9.1.4.2.c.2 and stated that it was used " exclusively for handling spent fuel assemblies within the fuel storage area..." in order to use the SFP Bridge and Holst to assist in the gate seat replacement a 50.59 safety evaluation was necessary.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to determine the acceptability of using the SFP Bridge and Holst to assist in the SFP gate seal replacement. The safety evaluation determined the weight of the SFP gate seal assembly to be approximately 1100 pounds including the rigging. The nearest SFP rack containing spent fuel is _15 feet away from the SFP gate. The rated load limit of the SFP Bridge and Holst is 4000 pounds and there is an administrative limit of 2100 pounds with an overload cutout of 2500 pounds. The SFP gate seal assembly was lifted and moved away from the direction of the SFP fuel storage racks. The lift was supervised 6nd performed by rigging and crane qualified mechanics and

- operators. The load was not carried over the spent fuel.

Considering these precautions and controls the safety evaluation determined that it was appropriate to use the SFP Bridge and Holst to assist in the gate seat replacement. A UFS_AR change to Section 0.1.4.c.2 allowing the use of the SFP Bridge and Holst for this and'similar SFP related tasks was approved. The safety evaluation concluded that no unreviewed safety question was created by this change.

14 of 46 '

i STAND ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER:

97 113 001 TITLE: Cooling Tower Painting and Spraying Activities Y

SUMMARY

DESC0lPTION AND PURPOSE:

In orde* to prevent corrosion and otherwise enhance preservation of material components in the Service Water Cooting Tower basin, painting activities were scheduled while the plant was at power. Platforms for workers to stand on and tenting material to prevent debris from entering the cooling tower besin were required. The platforms did not degrade the ability of the cooling ower to perform its safety related function, but the tenting material did. A plan was iniplemented to ensure the tenting material could be removed before the sprays and f ans in th-cooling tower were necessary.

l SAFETY EVALUATION SUMMAdY:

A 10CFR60.69 safety evaluation was completed to evaluate the effects the painting evolution would have or, the cooling tower with the plant at power. A safety relat9d calculation determined that it takes a minimum of 74 minutes following the switchover to the coolitig tower during a design basis accident before cooling tower sprays and fans are necessary. Procedural controls were established to ensure the fans and spray control switches were tagged out of service in the control room. Procedural controls were also established that directed operators to contact painting personnel and instruct them to remove the tenting material should the need arise to use the cooling tower. It was estirnated that it would take 30 minutes to remove the tenting material so that sprays and fans could be used.

The cooling tower remained operable throughout the painting evolutions. The safety evaluation concluded that no unreviewed safety questka was created by this change.

15 of 46

STAND ALONE 50.59 SAFETY EVALUATION: F. VALUATION NUMBER: 97 037 001 TITLE: Revised Safety Evaluation for the HP Contractor Control Point and Weather Protection Walkway

SUMMARY

DESCRIPTION AND PURPOSE:

An alternative facility was necessary to support increased Radiological Controlled Area (RCA) access and egress during outages, in order to provide this facility, a Health Physics Contractor Control Point (HPCCPT) and Weather ProtectL Walkway (WPW) were designed and built. The HPCCPT contained 500 pair of freshly laundered Protective Clothing (PCs).

The HPCCPT was connected to the WPW and the WPW was fastened directly to the permanent building structure. The WPW was f astened to the permanent building structures using flashing and caulking to make it weather tight.

SAFETY EVALUATION

SUMMARY

All design and structural requirements for the HPCCPT and WPW that are necessary to ensure that the two facilities will not impact any permanent safety related structures and components were followed. The storage of 500 sets of PCs was evaluated to ensure any offsite release caused by a fire involving the potentially, slightly contaminated, PCs did not exceed Technical Specification limits. A fire detection cystem is part of the facility's design and a roving fire patrol monitor examined the areas outside the f acilities on a routine basis as long as the facilities were in use. Contamination surveys were conducted daily and radiation surveys were conducted weekly to ensure that the quantitles of radionuclides present did not exceed the estimated maximums. The safety evaluation cuncluded that no unreviewed safety question was created by this change, STAND ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER: 96 121-01 TITLE: Temporary Covering of EFW Sump Building Floor Drains

SUMMARY

DESCRIPTION AND PURPOSE:

During painting activities in the Emergency Feedwater (EFW) Pumphouse, three 4 inch floor drains were taped over and thus unable to provide drainage. The drains were taped over and the Control Room was notified as required in the procedure that controlled the painting activitiu. The effect the covered floor drains had on the design basis of the drains and the potential effect on equipment operability was not evaluated. A 10 CFR 50.59 applicability review had been completed for the painting procedure and it was determined at the time of the review that a 10CFR50.59 evaluation was not necessary. The covering of the EFW pumphouse floor drains was determined to constitute a change to the facility as described in the UFSAR t'ecause UFSAR Section 9.3.3.2.b.8 indicated the EFW pumphouse contains five,4 inch drains and two,2,5-inch floor drains.

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STAND ALONE 50.59 SAFETY EVALUATIONt EVAL NUMBER: 9612101 (Continued)

TITLE: Temporary Covering of EFW Sump Euiluing Floor Drains SAFE's ? EVALUATION

SUMMARY

A 10;FR60.59 safety evaluation was prepared to evaluate the covering of the EFW pump iouse (. eins. The safety evaluuion described the diameter of the different floor drains, connect 7g drainage piping and the ultimate discharge point, and the c!! water separt tor vault. The oil water separator vault processing capability is 75 GPM. The inlet valve si throttled tilimit its input to that amount. Thus the flow from the EFW pumphouse is limited to 75 GPl 4. The EFW pumpnouse floor drains all flow into a 4 inch common drain line.

The 4 inch d ain line and the 75 GPM maximum oil water separator vault limit restricts the effectiveness of the EFW pumphouse floor drains. The safety evaluation concluded that even with three of the 4 inch drains taped over, the EFW pumphouse drainage capability remained unchanged from the design 1. isis maximum of 75 GPM. The safety evaluation concluded that no unreviewed safety question was created by this change.

STAND ALONE 50.69 SAFETY EVALUATION: EVALUATION NUMBER: 96TPST03 TITLE: Chemical Cleaning of GSC System

SUMMARY

DESCRIPTION AND PURPOSE:

Corrosion of copper components it' the Generator Stator Coolant System (GSC) occurred in January of 1996. The corrosion wes caused by a change in the oxygen concentration of the GSC system water. The corrosion caused blockage in the GSC system and resulted in reduced GSC flow. In order to restore the GSC system flow to normal, it was necessary to chemically clean the GSC system. A contractor with successfulindustry experience in the on line chemical cleaning of GSC systems dt ed the GSC clean up. During the clean up process, it was necessary to temporarily increase the GSC system conductivity alarm because the addition of the cleaning chemicals would raise the conductivity of the system above the value of the normal alarm setpoint.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was prepared to support the required temporary setpoint change. The safety. evaluation was necessary because UFSAR Sections 8.2.1.3 and 10.2.2.2 describe the GSC system as an oxygen saturated, de lonized cooling water system vented to atmcsphere. The chemical cleaning evolution increased the conductivity of the GSC system above the limits of de-ionized _ water, and thus the change to the conductivity alarm was necessary. None of the accidents previously evaluated in the UFSAR were affected by the change and no new potential accidents were created by the change. The conductivity was kept below the manuf acturer's maximum limit for conductivity in the system and the manufacturer agreed with and supported the chemical cleaning process. The safety evaluation concluded that no unreviewed safety question was created by the chemical cleaning evolution, 17 of 46 1

STAND ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER: 97 WOO 1989 TITLE: ilse of Strongback Device on 1 MS V296 SI'MMAFiY DESCRIPTION AND PURPOSE:

A valve packing gland leak on MS V 296 B" Steam Generator vent valve could not be stopped because the valve packing gland bolt was damaged. A strongback device was manufactured that could be used to hold the packing glane in place while the bolt was d

replaced.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was completed for the proposed mairm.1ance activity. The installation of a strongback device on MS V296 alters the method of performing the function of sealing the valve packing gland and thus constitutes a change in the facility as describet in the llFSAR. The 10CFR50.59 safety evaluation described the weight of the strongback and how it would be attached to the valve. The increased seismic bnd deadweight stress as a result of the strongback was calculated and compared to the corresponding allowable stresses and found acceptable. The worst cese scenario of a failure of the valve and the resultant 1 Inch steam leak were evaluated. The UFSAR Chapter 15 accident analysis for Steam System Piping Failure was reviewed and found to be bounding for the worst case scenario. The safety evaluation concluded that no unreviewed safety question was created by this change.

STAND ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER:

96-052 001 TITLE: Service Water Pump Change Out

SUMMARY

DESCRIPTION AND PURPOSE:

North Atlantic has found it necessary on occasion to replace a Service Water (SW) pump.

- The change out of a SW pump is a maintenance activity that requires lif ting of heavy loads in the vicinity of the safety-restated structures, systems and components, in order to safely perform this activity at power, the remaining operable train of the system and the pressure boundary discharge isolation valve and downstream discharge piping of the inoperable train must be maintained within the system design basis.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this DCR. The lifting of the pump components and roof hatch covers fall under the requirements of NUREG 0612 administered through the North Atlantic Lifting Systems Manual. For this lift, the crane, slings, cables, sha':kles and other lif ting equipment are sized for a minimum of two times the lif ted load, thereby satisfying the requirements of NUREG-0612 and Station commitments regarding the use of mobile cranes in NUREG 0612 areas. Further, the Pump 18 cf 46

STAP.D ALONE 50.59 SAFETY EVALUATION: EVALUATION NUMBER:

96 052 001 l

(Continued)

TITLE: Service Water Pump Change Out House roof structure has been evaluated for the effects of a heavy load drop. The evaluatloa demonstrated that the roof structure has adequate capacity to resist the impact of a heavy load and will proter:t the System. Roof hatch covers over the protected train are lef t in place as a barrier against a dropped load, in addition the potential for damage to a SW pump discharge isolation valve and downstream piping by a dropped load within the bv1 ding has been evaluated.

That assessment found that the application of the NUREG 0612 defense in depth approach, coupled with the physical arrangement, and shielding structures (i.e., pipe support) prevent damage to the isolation valve from a postulated load release. The pump discharge piping bellows, reducer, and check valve between the pump nozzle and the first pipe support are temporarily supported in a manner that prevents transmission of additional seismic or dynamic impact loads to the operable piping and discharge isolation valve. The safety evaluation concluded that no unreviewed safety question was created by this change.

STAND ALONE 60.59 SAFETY EVALUATION: EVALUATION NUMBER:

97 086 001

SUMMARY

DESCRIPTION AND PURPOSE:

This DCR removed the Slow Under Voltage Transfer (SUVT) feature from the Inverter 1 EDE-CP 1E static transfer switch. This DCR was to correct a condition where the SUVT could potentially swap the inverter loads from the regulated inverter output to the unregulated bypass output unnecessarily during certain operating conditions. Although the purpose of the SUVT is to tranrier the vitalloads to the bypass supply in order to preserve and sustain the UPS output, it is undesirable for a transfer to occur during a load application that could jeopardize the supply to the safety related loads.

SAFETY EVALUATION

SUMMARY

The safety r uation concluded that it was justifiable to remove the SUVT feature from the transfer schame. The basis for this justification was that there are other transfer circuits that will ensure the UPS output is transferred when appropriate and that the undesirable transfers must be eliminated. The safety evaluation concluded that the fai!ure mode analysis was unchanged as a result of the SUVT elimination and that the change will improve transfer switch operation as well as improve the functionality of the static transfer switch as it pertains to the design basis.

The existing accident analysis envelopes conditions if any part of the UPS or static transfer swit:h fails thus the consequences of an accident as analyzed are unchanged. The safety evaluation concluded that the SUVT is an optional feature and is not required to meet the design basis of the system. The safety evaluation concluded that no unreviewed safety que:; tion was created by this partial DCR.

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COMMITMENT CHANGE REQUESTS:

COMMITMENT CHANGE:

CCR NUMBER 97-03

SUMMARY

DESCRIPTION AND PURPOSE:

CCR 97 03 changed North Atlantic's commitment to NRC Generic Letter (GL) 88-17, " Loss of Decay Heat Removal." GL 8817 required licensees to implement certain actions prior to-operating in a reduced inventory condition. _ One of the required actions was to ' maintain sufficient existing equipment in an operable or available status so as to mitigate loss of Decay Heat Removal or loss of Reactor Coolant System (RCS) inventory should they occur," A high head injection pump and one other system was required to satisfy the requirements of GL 8817, North Atlantic committed to maintaining a Centrifugal Charging Pump (CCP) available and a gravity driven flowpath from the Refueling Water Storage Tank (RWST) to the RCS prior to operating at reduced inventory conditluns. Analyses conducted in response to GL 8817 determined that in order to satisfy the RCS makeup requirements using the RWST gravity 2

flowpath, a larga (18 in ) vent in the RCS was ne::essary to preclude RCS pressurization that could preverd gravity feed from the RW! T.

The successes that other plants had in using a new technique of vacuum assisted RCS fill and vent at the conclusion of refueling outages caused North Atlantic to change its commitment to GL 88-17 in order to use the new fill aad vent method. Vacuum assisted RCS fill and vent requires that the RCS not be vented. This prevented relying solely on the RWST at the second source of RCS makeup as the commitment to GL 8817. The commitment to maintaining a second source of RCS makeup prior to operating at reduced inventory conditions was changed to a combination of gravity feedpath from the RWST with no large vent; Safety injection (SI) accumulators at 80-90 psig and keeping available a second high head CCP.

SAFETY EVALUATION

SUMMARY

A 10 CFR 50.59 evaluation was performed for this commitment change. The safety evaluation referenced a safety related calculation that was done to predict the time to boiling following a loss of Decay Hect Removal while in a reduced inventory condition. The calculation estimated the time as approximately 29 minutes.

The safety evaluation

. concluded that as long as gravity feed from the RWST to the RCS was initiated within the time to boil, the mass addition to the RCS from the RWST would result in a large increase in RCS inventory and reduced inventory conditions would no longer exist.

With the RCS no longer in a reduced inventory con'dition and'with either a small vent or, the Technical Specification manual vent, or an open Power Operated Relief Valve (PORV),

the calculation shows that even with a continued loss of decay heat removal, core uncovery would not occur for approximately 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />.

20 of 46

COMMITMENT CHANGE: -

CCR NUMBER 97 03 (Continued)

The use'of Sl accumulators provides a means of RCS inventory addition at a pressure

-higher than that allowed by' gravity feed from the RWST. The accumulator _ pressure is limited by Technical Specification in this mode to less than 100 psig. The accumulators are sized such that the mass eddition to the RCS from the four accumulators would result in approximately the same time to core uncovery with a continue loss of Decay Heat Removal as that' afforded by the RWST-gravity flow rnakeup to the RCS, 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. In addition to'the use _of the RWST gravity feed flow path and the pressurized accumulators a-second CCP will be available for RCS makeup prior to operating in a reduced RCS inventory condition. Only a single clearging pump is allowed to be operable in mode 5 and 6 but a second CCP will be kept available. If the first CCP would become inoperable, the second CCP would quickly be made operable to provide RCS inventory makeup.

~

The 50.59 safety evaluation for CCR 97-03 concluded that the ability to rapidly restore RCS inventory using the RWST gravity flowpath, the makeup capability of the Si accumulators, and the availability of the second CCP provide a level of protection that is

. commensurate with the requirements of Generic Letter 8817.

The safety evduation concluded that no unreviewed safety question was created by this CCR.

COMMITMENT CHANGE:

CCR NUMBER 97-04

SUMMARY

DESCRIPTION AND PURPOSE:

CCR 97 04 revises a commitment documented in NRC Inspection Report 96-04 to inspect Unit Substation EDE US-64 at the next opportunity. The inspection was to investigate an unusual odor and potential increased noise level coming from the transformer.

The inspection was to take place during the fifth refueling outage (ORO5) since the Technical Specification (TS) allowed outage time for taking US-64 out of service was eight hours.

Subsequently, North Atlantic received a license amendment that increased the subject allowed outage time to seven days. Due to workload during OROS and since the transformer had been performing reliably since it was installed in November 1995, it became desirable to postpone the inspection, it was now possible to perform the inspection while at power as a result of the TS change. CCR 97-04 justified delaying this inspection until after the refueling outage. The inspection was sucessfully,ompleted during System Week 13, the week of August 24,1997.

-- SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this CCR. The safety evaluation concluded that design and safety evaluations dependent upon these provisions remain valid and no design or safety margins have-been reduced.- The safety evaluation concluded-that no unreviewed safety question was created by this CCR.

21 of 46

_-=.--.

~ _ -.

-1

- PROCEDURES, PROCEDURE REVISIONS AND PROCEDURE CHANGES:

The following procedures, procedure revisions and/or procedure changes were approved-during-the period covered by this 10 CFR 50.59 Report.

A safety _ evaluation was

.j performed -for each procedure, procedure revision -or procedure change.

Each safety -

evaluation con _cluded that the procedure, procedure revision and/or procedure change listed-below is available on site for review. Note: revision No. "00" is the initial issue of the procedure.

hocedura_

'Changa Revision Iltle OS 96-1-1 00

_00 EFW Dedicated Operator for ACR 96-1378

SUMMARY

DESCRIPTION AND PURPOSE:

OS 9611 was written to detail the responsibilities of a dedicated operator established to manually start Emergency Feed Water (EFW) flow during certain UFSAR Chapter 15 analyzed accidents. A scenario was identified during which automatic actuation of EFW would not occur if a steamline or feedline break occurred in the "A" Steam Generator (SG)

- coincident with a f ailure of the "B" Train Solid State Protection System (SSPS).

An operability determination for the potential scenario was performed which concluded that t

_ the EFW system was operable and operator action could be used instead of automatic EFW.

actuation.

SAFETY EVAL.UATION

SUMMARY

A 10 CFR 50.59 safety evaluation was completed in conjunction with OS 96-1-1. The safety evaluation concluded that the procedure made a change to the facility as described in the_UFSAR and made a change to procedures described in the UFSAR. The operability determination completed for the potential scenario concluded that the UFSAR Chapter 15 a.:.cident analyses acceptance criteria would be satisfied assuming EFW flow was manually initiated within ten minutes following the "A" S3 fault. The safety evaluation conclude that one of the two operators assigned to the control board are would be able to manually start the EFW flow within the required ten minute time frame.

Taking credit for operator action instead of automatic EFW initiation for this scenario was based on the fact that the operators are extensively trained on the emergency procedures used in this scenario. Additionally, status indication of EFW flow is readily available on the main control board, the operator action is simple and sufficiently proceduralized, adequate input signals are readily available to mandate manual EFW pump start and verification of EFW pump status is performed as part of the reactor trip response procedures which would be used by the operators in this scenario.

The safety evaluation concluded that no unreviewed safety question was created by this proced/e change.

22 of 46

- Ikoceduta.

Change Bevision Iltle SITR:

00 06 inservice Testing Reference -

SUMMARY

DESCRIPTION AND PURPOSE:

The Inservice Testing Reference (SITR) details the methods and criteria used by Seabrook Station to comply with the inservice testing of pumps and valves in compliance with Technical Specification 4.0.5 and 10 CFR 50.55a(fl. The inservice testing program was prepared in accordance with the rules of the ASME Boiler and Pressure Vessel Code,Section XI. The section of the SITR that describes relief valve tuting states that the testing is done in accordance with' ASME Section XI IWV 3510 that endorses ASME PTC 25.3 1976.- This change to the SITR clarified the exceptions to ASME PCT 25.3 1976 taken by North Atlantic because of the nature of our testing program.

SAFETY EVAt.UATION

SUMMARY

A 10CFR50.59 safety evaluation was completed for this change to the SITR. ASME PCT 25.31976 gives very prescriptive responsibilities to the " person supervising the test." The safety evaluation points out that the requirements in PCT 25.3 were written with relief valve manufacturers in mind and with the understanding that minimal procedures existed at the time.

The goal of PCT 25.3 was to ensure that a responsible person with an appropriate educational background was directly supervising relief valve testing. North Atlantic does not have one indisidual who is responsible for direct!v supervising all relief valve testing. _ North Atlantic ensures that relief valve testing is performed acceptably by conducting the testing in accordance with station programs and procedures under the Work Control and Operational Quality Assurance Programs. This change to the SITR clarifies the roles of individuals and groups in the North Atlantic organization with respect to relief valve testing.

This change to the SITR deals with a clarification in respcasibilities for testing of relief valves and does not physically change the plant. The SITR assures that the relief valves will be tested and set to the values required by the design bases while maintaining the intent of the requirements of PCT 25.3.

North Atlantic maintains the same level of assurance that the relief valves will be tested and set correctly. The safety evaluation concluded that no unreviewed safety question was created by this change.

23 of 46

L Panceduta. :

Change Bevisiott Iltle.

SSTR:

00--

46 Technical Requirements Manual SLTR-00_

01 Leakage Test Reference Manual-

SUMMARY

DESCRIPTION AND PURPOSE:

On February 4,1992 the NRC published the agency's planned actions and initiatives to eliminate requirements that are marginal to safety and yet impose significant regulatory burden on licensees. The NRC determined that the detailed and prescriptive technical requirements contained in some regulations could be improved and replaced with performance based requirements and supporting regulatory guides. One example of the NRC change in philosophy was a modification to 10 CFR 50 Appendix J, containment leakage and testing requirements.

' A performance based containment leakage testing option, Option B of 10 CFR 50 Appendix

- J, became effective on October 26,1995. This option allows the use of a reduced testing frequency for primary containment systems and components based on performance history.

North Atlantic applied for and was granted an amendment to Seabrook Station's operating liconse, Amendment 49, on February 24,1997. As a result of the change to Option B, Appendix J testing, changes were necessary to the Seabrook Station Technical Requirements (SSTR) and the Seabrook Leakage Test Reference (SLTR).

SAFETY EVALUATION

SUMMARY

The NRC reviewed the potential impacts on safety associated with the irr c ementation of Option B and documented the impact in NUREG-1493, " Performance-Dased Containment Leak-Test Program." The NRC concluded that the impact of reducing tha frequency of containment leakago tests were imperceptible. As part of the Seabrook Station Appendix J Scheduler Exemption for. the fourth refueling outage, an extensive evaluation of the Appendix J test program and the associated companion programs was performed. This evaluation concluded that extending the intervals for the containment leakage testing at Seabrook was justified for the same reasons outlined in the NRC review. The change to

-Option B containment leakage testing in accordance with Regulatory Guide 1.163 will provide adequate assurance that the containment will perform its design function and will have minimal impact on public health and safety or radiological releases.

The safety evaluation concluded that no unreviewed safety question was created by this change.

4 24 of 46

i Erocedute Chansa Beklalon Iltle OS1013.03/04 11/12 09/09 RHR FLUSH TRAIN AlTRAIN 8

SUMMARY

DESCRIPTION AND PURPOSE:

The purpose of these procedures is to provide a flowpath that can be used at power to reduce dose rates in the mechanical penetration area and the equipment vaults. This flowpath must provide adequate flow to maintain occupational doses ALARA and maintain both trains of RHR operable, in order to consider the RHR system operable during this evolution, any flowpath used must be automatically isolated upon a safety injection signal or be limited in-flow such that the diversion from ECCS injection does not reduce the injection f.ow to a value below that assumed in the LOCA analysis. These procedures were revised to document that an acceptable flowpath has been identified through RH V18 (V19) and HCV-128 to the letdown subsystem, through the demineralizers, and returning to the RHR system via CS-V828 (829). Also, flowpaths through RH-V16 (V17) and the Si accumulator test lines will be used.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed for these procedures. The determination of the' safety evaluation concluded that the procedures do make changes to the facility as described in the UFSAR and do make changes in procedures as described in the UFSAR.

Opening RH V18 (19) and throttling the flow to 60-75 gpm via HCV-128 ensures that the required injection flow is available to the RCS following a LOCA. The other valves that are opened are automatically isolated on accident signals and thereby preserve the required injection flow. The safety evaluation concluded that no unreviewed safety questions were created by these prucedure changes.

25 of 46

Etocedurs Change

- Bavision Illla ES1801.009 00 00 Emergency Feedwater Turbine Overspeed Test Using Auxiliary Steam

SUMMARY

DESCRIPTION AND PURPOSE:

This procedure provided a means to cross-connect auxiliary steam to the EFW turbine for purposes of performing the overspeed_ trip test earlier in the outage schedule. The recommendation is supported by the vendor in the EPRI NP6909 Research Project 2814-14, Terry Turbine Controls Guide. The EPRI document stated "It is recommended that these test be done on low pressure, 100-250 psig, steam to allow easier control and minimize the chance of runaway." The use of auxiliary steam allows the overspeed test to be done when the EFW turbine is available for testing as opposed to the end of the outage.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed on this procedure. The safety evaluation determined that the procedure did involve changes to the facility as described in the UFSAR (Section 3.6, Appendix 3A, Appendix 31, 6.8, 10.3.1, 10.3.5, 10.4.3, 10.4.11, 13.5,14.2, and 15.0). This test performs the routine exercising and testing to ensure the operability of the device when it is needed. These tests are done using approved station procedures that follow vendor and industry guidelines. These tests are done when the system is not required to be operable, and in an operational mode in which no adverse effect on the safe operation of the plant can exist. The safety evaluation concluded that no-unreviewed safety question was created by this procedure change.

26 of 46

~

Etoceduta

-Change Baylainn Iltle MS0504.01_'

03 05_

CRDM - Missile Shield Platform ' Installation and-and Removal MS0504.25

SUMMARY

_ DESCRIPTION AND PUPNSE:

Procedures MS0504.01/,25 were revised to allow the upper geared limit switch for the Polar Gantry Crane (PGC) to be bypassed during lif ting of the CRDM Missile Shield platform and related activities as described in the subject procedures. The change also provided for compensatory actions to be taken while the geared limit switch is bypassed. Specifically, steps were added to p;sition an individual who is in direct contact with the PGC operator to observe the PGC main hook while the upper geared limit switch is " jumped out", and that will reduce the crane hoisting speed The observer is responsible for directing the PGC

-operator to immediately stop hoisting operations if main hook hoist block contact with the paddle limit switch is imminent. The reduced hoist speed provides more margin for the-observer and operator to react. The paddle limit switch will be active and will provide the necessary degree of redundant protection against excessive hook travel in the upward direction.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed for these procedures. The safety evaluation determined that the procedures did make changes to the facility as described in the UFSAR (Sections 9,1.4, 9,1.4.2, and Chapter 13.5). The purpose of the limit switch is to prevent "two blocking" that can result in damage to the wire hoisting rope. The PGC will continue to. function as designed and will not be subject to any unanticipated failures.

Reducing hoisting speed while providing an individual to observe the main hook position, in conjunction with the existing paddle limit switch will provide assurance that "two blocking" will not occur. The safety evaluation concluded that no unreviewed safety question was created by this procedure change.

27 of 46

Procedute Change Revlalon Iltle.

OX1405.07 09 06 Safety injection Quarterly and 18 Month Pump Flow and Valve Test

SUMMARY

DESCRIPTION AND PURPOSE:

This' procedure change, requiring the use of a-dedicated operator, is a result of corrective actions associated with ACR 95 238. ACR 95 238 was generated to resolve an industry review item that determined that closing CBS-V49, Train A, SI pump suction isolation valve, isolated the flow path from Train B RHR pump to the Train B CCP during the ECCS recirculation mode of core cooling. Surveillance procedure OX1405.07 was revised to require a dedicated operator at valve CBS-V49 during the valve stroke test. Use of a dedicated operator is proceduralized and does not replace an automatic function.

SAFETY EVALUATION

SUMMARY

1 A 10CFR50.59 safety evaluation was performed for this-procedure change. The determination of the safety evaluation applicability concluded that the procedure made changes to the facility as described in the UFSAR (Figure 6.2 75) and does make changes in procedures as described in the UFSAR (Table 6.310). The safety evaluation concluded that the combination of the dedicated operator, the e: vironment, the short action time and the time available provide reasonable assurance that the safety function is met. The safety evaluation concluded that no unreviewed safety question was created by this procedure change.

l l

l I

4 28 of 46

- haceduta

-Chansa Bevision Ihla LS0565.20 00 00

. Bus 5 Outage Temporary Power Installation

SUMMARY

DESCRIPTION AND PURPOSE:

'The purpose of this procedure is to denote the restoration of criticalloads on Bus 5. During regularly _ scheduled maintenance periods it is necessary to de energize the 4.16 KV Bus E5 for cleaning, inspection, and general preventative maintenance. During such outages there are critical loads that are highly desirable to remain available, implementation of all or part of this procedure shall only be performed when the station is in Mode 5,6, or defueled and

-the operable AC electrical power source is the B train. Additionally, this procedure shall be restored prior to declaring the A train operable, in all cases where temporary power cables are being installed per this procedure it is required that they be routed external to the permanent raceway systems.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed on this procedure. The determination of the safety evaluation applicability concluded that the procedure did make changes to the facility as described in the UFSAR (Sections 1.2, 7.1, 7.2, 7.3, 8.3,- 9.5, 10.4.8, and 13.5). Prior to implementing this procedure the reactor shall be in Mode 5, 6, or defueled and shall be restored prior to declaring the A train electrical distribution system operable.

All modifications of this procedure will not impact the operability of the B train electrical system. The safety evaluation concluded that the procedure did not adversely affect the design basis or wnction of safety related structures, systems, or components. The safety evaluation concluded that no unreviewed safety question was created by this procedure change.

29 of 46

Pmcadura Change Bavision Iltle OS 9714 00 00 Temporary Modification 97-02 Dedicated Operator

SUMMARY

DESCRIPTION AND PURPOSE:

This procedure delineates the responsibility and acceptability of the use of a dedicated operator to provide continuous monitoring of the Spent Fuel Pool during implementation, operation, and removal of the Temporary Modification 97-02. The dedicated operator will provide local control of the temporary cooling system and Spent Fuel system status monitoring (pool temperature, level, etc.) during the period of time that the SF pumps are isolated, while the temporary cooling system is operating, and during restoration of the SF cooling pumps. During the temporary cooling pump operation if temporary backup power source is required, the dedicated operator will perform actions necessary to place the

~ temporary cooling system in a shutdown condition and align alternate power.

- SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed for this procedure. The determination of the safety evaluation applicability concluded that the procedure made changes to the facility and involved tests not described in the UFSAR (Section 9.1.3, 15.7.2, 15.7.3, 15.7.4~ and 15.7.5). The Spent Fuel Pool Cooling System will still provide the necessary water inventory and cooling effect for the stored spent fuel assemblies for the short duration when forced SFP cooling is not in service. The safety evaluation concluded that the procedure did not involve an unreviewed safety question.

Procedute Change Bevision Iltle ON1042.01 12 07 Operation of the Compressed Air System

SUMMARY

DESCRIPTION AND PURPOSE:

Revision 7, change 12 of procedure ON1042.01 " Operation of the Compressed Air System", provides steps for aligning portable compressors to supplement system capacity duritig periods of high demand or to support maintenance-of the permanent equipment.

- The. ortable compressor connections (SA V1165/SA-V288) will provide either Service Air (SA), Instrument Air (IA), or SA and IA. Each of the headers has local and remote pressure iridication and low pressure alarms. Portable air compressors will be oil free and will be operated within the design limitations of the compressed air system.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed on this procedure. The determination of the safety evaluation concluded that the procedure did make changes to the facility as described in the UFSAR (Section 9.3.1). The alternative flow paths described will augment the system capacity, availability, and reliability.

The safety evaluation concluded the revision was an enhancement and did not-adversely affect the design basis or function of safety related structures, systems, or components. The safety evaluation concluded that no unreviewed safety. question was created by this procedure change.

30 of 46

PROCEDURE / PROGRAM UFSAR CHANGE:

TITLE: Process Control Program for Spent Resin and Filters

SUMMARY

DESCRIPTION AND PURPOSE:

The Seabrook Station Process Control Program (PCP) describes the administrative and technical controls on the radioactive waste processing systems. These systems provide assurances that the NRC regulations and disposal site license requirements involving waste from structural stability and free standing water are met. The PCP has been revised to include the processing (dewatering) of spent resin beads and process filter cartridges within the Waste Process Building (WPB) and under the adherence of the applicable Rad Waste Department procedures.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed on this proposed process. The safety evaluation applicability concluded that the process did involve changes to the facility as described in the UFSAR and to procedures as described in the UFSAR. The following sections of the UFSAR were reviewed: 11.1,11.2,11.3, and 11.4. The PCP required

- revision to state that spent filter cartridges will typicasy be processed by the dewatering methods already in place for bead resins. Seabrook Station maintains the alternate option of processing spent filters through encapsulation and solidification. The conclusion of be safety evaluation is that no unreviewed safety question was created by this program change.

TEMPORARY MODIFICATIONS (TMODs)

The foibwing temporary modifications (TMODs) were installed during the period covered by this 10 CFR 50.59 Report. In each case, a safety evaluation was performed. Each safety evaluation concluded that the TMOD did not involve an unreviewed safety question.

Summaries of these TMODs and their safety evaluations are provided below.

31 of 46

i TEMPORARY MODIFICATION (TMOD) NUMBER:

95 0036, Rev,02 TITLE: Lif ted Leads for PCCW Flow Switches to CAH FN-1 A through 1F

SUMMARY

DESCRIPTION AND PURDOSE:

This TMOD is to provide the means to operate the Containment Structure Cooling Subsystem, by removing.the PCCW flow switches from the control circuit. This will improve the ability to maintain a constant temperature with the equipment hatch open,

- thereby avoiding a potentiallow temperature problem.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed on this TMOD. The determination of the safety evaluation applicability concluded that the TMOD did make changes to the facility as described in the UFSAR (Section 9.4.5.1b). The system is not required to mitigate an accident and has no design bases. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

TEMPORARY MODIFICATION (TMOD) NUMBER:

96-016 TITLE:

Disabling Reactor Coolant Pump RC-P 1D Horizontal Shaf t Vibration Probe

SUMMARY

DESCRIPTION AND PURPOSE:

This TMOD disabled one of the two horizontal shaft vibration probes on the "D" Reactor Coolant Pump (RCP). One of the two horizontal shaft vibration probes was determined to be operating erratically. As a result of the erratic operation, numerous spurious alarms were being generated in the Control Room and distracting the operators. The redundant horizontal shaft vibration probe accurately displayed in the Control Room and provided full monitoring and alarm capability of the "D" RCP.

SAFETY EVALUATION

SUMMARY

The 10 CFR 50.59 safety evaluation performed in support of this TMOD determined that this TMOD made a change to the facility as described in the UFSAR. UFSAR Section 5.4.1.2.b describes two redundant horizontal shaft vibration probes and two frame vibration probes that monitor each RCP. The safety evaluation concluded that the three remaining vibration probes allowed for adequate RCP vibration monitoring. This change did not affect the ability of the RCP to perform it's intended function. The safety evaluation concluded tuat no unreviewed safety question was created by th's TMOD.

32 of 46

TEMPORARY MODIFICATION (TMOD) NUMBER:

96 019 TITLE: Temporary Power for Admin Bldg Lightir.g

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modification supplied 480v power from an Exeter-Hampton supplied load center (LC15T) to power panels 166 A/B and 167, which are normally supplied from Unit Sub (US) 16. US 16 is shown in the UFSAR, Figure 8.3.1. All loads from US 16 are non safety.

This temporary modification does not make changes in procedures nor does it constitute a test / experiment described in the UFSAR.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this TMOD. The description of US 16 in the UFSAR notes that all loads are non-vital and therefore the temporary modification does not increase the probability of an accident /maifunction of a different type than previously evaluated in the UFSAR. There is no basis in the Technical Specifications relative to the non safety related loads powered from US 16, therefore, the margin of safety as defined in any basis is not reduced. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

TEMPORARY MODIFICATION (TMOD) NUMBER:

96-024 TITLE:

Mt Plant Computer System (MPCS) A Points for US 16 Current and Voltage

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modification provided analog transducer signals to the MPCS. This temporary modification added a current transducer to the local ammeter and utilized the existing voltage transducer on US 16 to provide analog inputs to the MPCS.

SAFETY EVALUATION

SUMMARY

US 16 and its associated metering are shown in the UFSAR, Figure 8.3.1, although there is no specific mention of US 16 or its loads. The temporary modification neither makes changes to procedures described in the UFSAR nor does it constitute a new test or experiment._The UFSAR does not reference US 16 or its attendant loads with regard to accident or malfunction impact. US 16 supplies non-vital loads only and has no basis in Technical Specifications, therefore, the margin of safety as defined in the basis is not reduced The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

33 of 46

TEMPORARY MODIFICATION (TMOD) NUMBER:

96-032, Ra 03 TITLE: Forced Cooling of Steam Generator Wet Lay up Line

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modificatio_n removed pipe. insulation and provided forced cooling of _the exposed SG Wet Layup line by temporary fans. The leakage of hot feedwater past the seat of' FW V153 (used to establish steam generator recirculation flow path) _ into the EFW

' header could potentially result in damage to piping and/or stearn binding of EFW-as described in Information Notico 84-06 and IE Bulletin 85 01. This TMOD will cool the piping until permanent valve repairs can be made to FW V153. This ection in conjunction with the operator and technical support monitoring programs will ensure continued protection of the EFW pumps from steam binding.

SAFETY EVALUATION

SUMMARY

- A 10CFR50.59 evaluation was performed for this TMOD.

The Feedwater System is-

- described in Section 10,4.7 and the EFW system is described in Section 6.8 of the UFSAR.

A review of these sections shows no procedural requirements or test / experiments affected by this temporary modification, The TMOD is non-intrusive and does not alter the system function in any way. Therefore the probability or consequences of an accident or malfunction are unaffected. This TMOD does not affect any Technical Specifications or their design basis. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

34 of 46

TEMPORARY MODIFICATION (TMOD) NUMBER:

96-034

. TITLE: Spent Fuel Bridge Temporary Modification

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modification teoriented the spent fuel bridge to allow the placement of spent fuel in sixty open fuel cells, located in the west end of the pool. This allows for

" Full Core Discharge" capability through the sixth refueling outage. The orientation will be performed using the Spent Fuel Cask Handling Crane.

SAFETY EVALUATION

SUMMARY

~ A 10CFR50.59 evaluation was performed for this TMOD. The Fuel Handling System is described in UFSAR Section 9.1.4 and the Spen * ' uel Bridge is shown on Figures 1.216, 1.218, and 1.2-21. The bridge restricted travel interlock at the gate to the cask loading area is described in UFSAR Section 9.1.4.3(d). The fuel transfer system lifting arm interlock with the spent fuel bridge and hoist is found in UFSAR Section 9.1.4.3.a.2(f). The temporary modification does make a change to the facility, how<

. it does not require changes to any procedures and it does not involve any tests or eN riments not described in the UFSAR. The requirements of NUREG-0612 do not apply. UFSAR, Chapter 15 Section 15.7.4.3.b discusses the radiological consequences of a fuel handling accident inside the Fuel Storage Building. Mitigation of the postulated accident is accomplished by the Fuel Storage Building ventilation system that is unaffected by this design change. The original-design of the spent fuel handling system remains intact therefore there is no increased probability or consequences to an accident or malfunction different from or already described in the UFSAR. There is no effect therefore no reduction in the margin of safety used as bases for the Technical Specifications. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

1 i

35 of 46 -

i

TEMPORARY MODIFICATION (TMOD) NUMBER:

96-035 TITLE:' Disabling of DG C 2A Oil Sump Level Switch (DG-LSL 9519A)

SUMMARY

DESCRIPTION AND PURPOSE:

This TMOD_ disabled the Diesel Generator (DG) air compressor tube oil sump level switch.

This switch is constantly in alarm due to a switch malfunction. This results in both the local annunciator 'and the common DG A trouble digital point being in alarm. With the common trouble alarm being in all the time, this masks other possible alarms from the Control Room. This TMOD will allow the other sensors to input to the common trouble alarm and restores the Control Room DG-A monitoring capability.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed for this TMOD The determination of the safety evaluation applicability concluded that the TMOD did make changes to the f acility as described in the UFSAR (Section 9.5.6.5) Compensatory measures will be taken to verify adequate sump level in the compressor. This TMOD does not adversely effect safety related systems, structures, or components. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

TEMPORARY MODIFICATION (TMOD) NUMBER:

96-039 TITLE: Temporary Power for MCC 272 During Bus 2 Breaker Maintenance

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modification supplied 480v power from an Exeter-Hampton supplied load center (LC 10 ckt.11) to non safety MCC 272. The load is shown in Figure 8.3.1 of the UFSAR, however none of the non safety MCC loads are detailed in the UFSAR A review of UFSAR Sections 8.3.1, 9.5.3, and Chaptor 13.5 indicates no changes in procedures or tests / experiments as described in the UFSAR.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 was performed for this TMOD. The description of MCC-272 in the UFSAR notes that all loads are non vital and therefore the temporary modification does not increase the probability of an accident / malfunction of a different type than previously evaluated in the UFSAR There is no basis in the Technical Specifications relative to the non safety loads powered from MCC-272. Therefore, the margin of safety as defined in any basis is not reduced. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

36 of 46 1

TEMPORARY MODIFICATION (TMOD) NUMBER:

97-001, Rev. 01 TITLE: Heated Enclosure for CBA AC 5A 8 5B

SUMMARY

DESCRIPTION AND PURFOSE:-

As a proactive measure,1while awaiting the implementation of DCR 94-025, this'TMOD provided a heated enclosure for Control Building Air Handling -(CBA) compressors. The TMOD heats the environment the compressor crankcases are subjected to thereby maintaining an oil viscosity that enables cold starts without concerns of e low pressure oil trip. This temporery heater is thermostatically controlled and periodically monitored by a fire watch and/or an operator to ensure design temperature for the equipment is maintained -

within specifications.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this TMOD The Control Room Area Ventilation Subsystem is a Safety Class 3, Seismic Category 1 and is designed to maintain the control room temperature and humidity within the limits specified on the Service Environment Chart (UFSAR Figure 3.11-1).

The CBA system is described in Section 9.4.1.2 of the UFSAR. The heated enclosures will only effect the area shown on Figure 9.419, the Control Building Air Handling Detail.

Operation of the' Control Room Area Ventilation Subsystem will not be affected by the installation of the heated enclosures. The enclosures have been reviewed for seismic and fire protection impact. The TMOD does not impact accidents postulated in Chapter 15.

These enclosures will be removed when ambient temperature is 250'F The addition of the heated enclosures will not increase the probability of an accident or malfunction of a different type than any previously evaluated in the UFSAR. The margin of safety as defined in the basis for Technical Specifications 3/4.7.6 will not be reduced. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

I 37 of 46

TEMPORARY MODIFICATION (TMOL NUMBER:.

97 002-

. TITLE: Temporary Modification for Spent Fuel Pool (SFP) Cooling _

SUMMARY

DESCRIPTION AND PURPOSE:

-This TMOD installed temporary pumps and attendant piping / hose to the downstream piping of SFP A heat exchanger to maintain SFP_ cooling when the normal cooling system was taken out of service for approximately 60 to 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> to support installation of a third.

SFP pump.

SAFETY EVALUATION

SUMMARY

A _10CFR50.59 evaluation was performed for this TMOD. This safety evaluation addressed temporary cooling to assura the SFP is maintained at an acceptable temperature. Standard -

Review Plan Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System, recommends that a normal maximum spent fuel pool temperature of 140'F not be exceeded. The cooling ensures SFP temperature is below that limit. The temporary cooling method is different from the Spent Fuel Pool Cooling System described in UFSAR Section 9.1.3. Since this is a temporary-condition a change to the UFSAR is not required. The temporary cooling configuration does not make changes to the procedures described in the UFSAR The temporary cooling configuration does involve tests or experiments not described in the UFSAR. Prior to removing the SFP Cooling System from service the temporary cooling configuration will be tested.

The-safety evaluation concluded that the TMOD did not increase the probability of an accident evaluated in the UFSAR since it did not affect the Standard Review Plan Section 9.1.3, UFSAR Sections 15.7.2, " Radioactive Waste System Leak or Failure",15.7.3,

" Postulated Radioactive Releases Due To Liquid Containing Tank Failures",15.7.4, " Fuel Handling Accident", and 15.7.5, "Spe.it Fuel Cask Drop Accident". Adequate measures and administrative controls were in place and the rnargin of safety of the SFP Cooling System was not reduced. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

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.. h -

k TEMPORARY MODIFICATION (TMOD) NUMBER:

97-004 TITLE: Containment Closure Fixtures'for Sludge Lance / Pressure Pulse and Eddy

Current Testing.

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modification removes flanges inside containment (noted in UFSAR Table 6.9-92) and installs outage " fixtures" for sludge lance. pressure pulse, and ECT during Mode 5 and Mode 6. This change does not compromt the containment building function,-

since the change is for Mode 5 and Mode 6. Considering a review of the UFSAR, Se:tions -

6.2.4, 6.2.6, and 15.7.4, the replacement of the blind flanges with the outage "fix ures" does not involve a change in procedures nor does it involve a test or experiment not described in the UFSAR.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 was performed for this TMOD.

The potentially affected postulated accident is that of a fuel handling accident inside containment (UFSAR, Section 15.7.4).

The basis for Technical Specification 3/4.9.4 is to ensure containment building closure such that a release of radioactive material within the containment will restrict leakage to the environment. The closure restrictions are sufficient to restrict radioactive material release from a dropped fuel element based on a lack of containmer.t pressurization while in the refueling mode. The temporary modification is in compliance with Technical Specifications 3/4.9.4. The temporary modification rnaintains a boundary that will prevent the exchange of the containment environment with the outside environment under postulated refueling accident conditions. The pronability of an increase nr different consequences of an accident or malfunction will not be impacted due to maintaining the containment closure requirements. The margin of safety will not be reduced. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

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TEMPORARY MODIFICATION (TMOD) NUMBER:

97 006 TITLE: Ternporary Cooling Water to SI P-6B and CS-P-2B

SUMMARY

DESCRIPTION AND PURPOSE:

During the fifth refueling outage, operation of Centrifugal Charging Pump (CCP), CS-P 28 and Safety injection (SI) Pump, SI P-6B was necessary to support Emergency Core Cooling System (ECCS) flow balance testing. The normal supply of cooling water to these pumps was unavailable during the ECCS testing. This TMOD establisted an alternate source of cooling water for these pumps during the test.

SAFETY EVALUATION

SUMMARY

A 10CFRSO.59 evaluation was ;npared to support this TMOD.

This TMOD makes changes to the f acility as described in t!a UFSAR (Sections 6.3, 9.2, and 9.3). During the period of time the ECCS flow balance test took place, the plar.t was in Mode 6, with the reactor core offloaded to the Spent Fuel Pool. The CCP and SI pump were net required to be operable per Technical Specifications and they were not required to perform a safety function during the ECCS test. The TMOD does not assure any required safety related function. The TMOD was removed and the normal cooling was realigned to the pumps before they were required to be operable or were required to perform a safety related function. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

TEMPORARY MODIFICATION (TMOD) NUMBER:

97-007 TITLE: Loss of dontrol Rod Drive Mechanism (CRDM) Fan (s) Computer Logic Change

SUMMARY

DESCRIPTION AND PURPOSE:

The' CRDM fans (4) provide cooling to the magnetic Jack CRDM coil stack winding assemblies. The failure of two CRDM fans had occurred which caused an alarm, (FS2S6 CRDM COOLING SYS FLOW HIGH/ LOW), on the main plant computer for less than three

-tans operating, This TMOD provides the computer logic changes requirod to initiate the alarm if either of the two remaining fans, CAH FN 2A or 2C should fail to operate. The CRDM cooling system has no safety related function and is non-seismic.

SAFETY EVALUATION

SUMMARY

A 10CFRSO.59 evaluation was performed for this TMOD. The determination nf safety evaluation applicability concluded that the TMOD made changes to the facility as described in the UFSAR. The safety evaluation included review of UFSAR Section 9.4.5.2 and determined that the TMOD would not adversely affect nuclear safety related structures, systems, or components. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

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TEMPORARY MODIFICATION (TMOD) NUMBER:

97-008 TITLE: Bypass CAP-V 1 and CAP-V-4 Solenoid Valves

SUMMARY

DESCRIPTION AND PURPOSE:

_This modification provides an alternate method of supplying air to CAP-V1 and V4 during the Bus 5 outage. The Bus 5 outage removed power to the solenoids that supply air to open CAP V1 and V4. These valves were opened to provide the filtered flow path for operation of the ore-entry purge supply and exhaust fans. Due to radioactive lodine levels in the containment building, the system remained in service. Using required f ttings and tubing the solenoid valves will be bypassed to supply instrument air to the actuators. This provided a continuous air supply to maintain the valves in an open position. This TMOD

.was installed while the plant was in Mode 6, after the vessel has been defueled. The TMOD was restored following the Bus 5 outage prior to reloading the core.

SAFETY EVALUATION

SUMMARY

1 A 10CFR50.59 safety evaluation was performed on this TMOD. The safety evaluation applicability de.ormined that the TMOD did involve a change in the facility as described in the UFSAR (Sections 0.2.4 and 9.4.5.2d). Technical Specifications apply in Modes 14 and during core alterations or the movement of irradiated fuel within containment. As stated above, this modification was implemented in Mode 6 after the core was off-loaded and was restored prior to core load. The safety evaluation concluded that no unreviewed. safety question was created by this TMOD.

TEMPORARY MODIFICATION (TMOD) NUMBER:

97-010 TITLE: Reactor Makeup Water (RMW) Cross Tie to Primary Side Demin Header

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modification provides a cross-tie between the RMW system and the DM system. The portion of the domineralized water header that extends from the turbine building will be isolated and drained to facilitate the addition of lines to support the condensate polisher installation. This modification provides an alternate water source for the domineralized header and does not interface with any safety related equipment or

- function thereof.

SAFETY EVALUAT90N

SUMMARY

A-10CFR50.59 evaluation was performed for this TMOD. The determination of the safety ovaluation applicability concluded that the TMOD made changus to the facility and to the procedures as defined in the UFSAR (Sections 9.2.3.1 and 9.2.7.1).

The safety evaluation concluded that the TMOD would not adversely affect nuclear safety related structures, systems, or components. The safety evaluation concluded that no unreviewed safety question we created by this TMOD.

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TEMPORARY MODIFICATION (TMOD) NUMBER:

97 011 TITLE: Provide Pump and Hose Hookups for Temporary DM Supply to CST and DWST

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modification provides temporary connections iri the Demineralized Water (DM), Condensate (CO), and Water Treatment (WT) systems to allow for the DM and CO storage tanks to be filled from a mobile water treatment unit (s) that will be available to supplement the permanent leased wates treatment facilities.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this TMOD. This ternporary modification does change the facility as described in the UFSAR (Sections, 9.2.3/9.2.6, Figures 9.2-7,9.2 8, and 10.4-6, and Table 9.2-11). This change involves the use of a separate (mobile) processing facility. The safety evaluation concluded that the TMOD would not adversely affect nuclear related structures, systems, or components. The safety evaluation concluded that the TMOD did not involve an unreviewed safety question.

TEMPORARY MODIFICATION (TMOD) NUMBER: 97 012 TITLE: Condensate Storage Tank (CST) Deox Recirculation

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modification installed a recirculation line on CS-TK 25 (CST). This line routes condensate to the domineralized water system to allow it to be processed through the DEOX/ Demineralized water processing facility (temporary facility utilized during Modes 4 6, and defueled).

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was performed for this TMOD. This temporary modification changed the facility as described in the UFSAR (Section 9.2.6, Figure 10.4.7 sheet 5). The temporary modification does not represent an increased probability of consequences or malfunctions as the modification was installed during Modes 4-6 when Technical Specifications for EFW and the CST are not applicable. The margin of safety as defined in the basis for any Technical Specification will not be reduced as the installation was performed when the CST inventory was not required. The safety evaluation concluded that

- no unreviewed safety question was created by this TMOD.

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TEMPORARY MODIFICATION (TMOD) NUMBER:

97-013 TITLE SCC Head Tank, CAS Pump Seal and AR Pump Seal Alternate Makeup

SUMMARY

OESCRIPTION AND PURPOSE:

This temporary modification (TMOD) replaced the normal demineralized water supply to the Secondary Component Cooling (SCC) water system head tank makeup, and the seal supply to the Chemical Addition System, Air Removal and Water Box Priming Pumps with water from the Condensate System (CO) during a Demineralized Water System (DM) outage.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed to support this TMOD. UFSAR Section 10.4.7 references DM water as the source of markup for the SCC head tank. The safety evaluation concluded that none of the systems involved in this TMOD are relied upon to mitigate the consequences of an accident. None of the systems are nuclear safety related.

The portions of the DM system being isolated and drained are not required for plant shutdown. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

TEMPORARY MODIFICATION (TMOD) NUMBER:

97-014 TITLE:

Temporary WT Waste Discharge to Unit 2 Forebay

SUMMARY

DESCRIPTION AND PURPOSE:

This temporary modification (TMOD? was nr. saary to allow continued use of the Water Treatment. System (WT) while the Circulating Water System (CW) was out of service during OROS. The waste water from the WT system is normally discharged through the CW system so a TMOD was necessary to allow the WT waste water to be stored in the Unit 2 CW forebay until the CW water system was returned to service.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was completed to support this TMOD. UFSAR Figure 9.2-8 depicts the WT system overview. While this overview does not show the piping detail affected by this TMOD, the TMOD does change the discharge point for the WT system as described in UFSAR Section 9.2.3. The safety evaluation determined that the TMOD did not affect. any accidents, consequences of accidents or - malfunctions of equipment important to safety. The WT system has no emergency function and is not required for safe shutdown. This TMOD had no effect on safety related equipment. The safety _

evaluation concluded that no unreviewed safety question was created by this TMOD.

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- TEMPORARY MODIFICATION (TMOD) NUMBER: 015 TITLE: Installation of Portable Gas Padiation Monitor at WRGM Low Range Sample Lines on Loss of Low Range Charinel P

SUMMARY

DESCRIPTION AND PURPOSE:

This TMOD installed a temporary radioactive gas monitor in place of the Wide Range _ Gas Monitor (WRGM), which monitors the RCA buildings ventilation exhaust-for radioactivity.

The WRGM is a Technical Specification monitor required to be in service at all times. The WRGM was removed-from service during refueling outage OROS.

The required grab sampling and analysis of the RCA building ventilation exhaust was completed per the Technical Specification action statement in addition, this TMOD put in place a temporary, in line, radioacJve gas monitor that provided continuous seal time monitoring and display of the RCA ventilation exnaust gas radioactivity level.

. SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was completed to support this TMOD. Section 12.3.4.2.c _ of the UFSAR describes the WRGM operation and components making up this effluent gas

monitor, in the UFSAR description of the WRGM the sampling provisions located downstream of the isokinetic nozzle are described. Since the TMOD was connected to the system at these sampling provisions, this TMOD was a change to the facility as described in the UFSAR.

The safety evaluation evaluated connecting the TMOD, using flexible tubing, to the sampling connections in the WRGM flow path. The TMOD was easily installed and removed so that the required grab samples could be collected. The TMOD did not pormanently alter or affect existing plant equipment o procedures. The temporary gas monitor did not affect any safety related equipment, did not pose any risk of increasing the possibility, probability or consequences of any accident or malfunction either new or previously identified in the UFSAR. The safety evaluation concluded that no unreviewed safety question was created by this TMOD, 44 of 46

TEMPORARY MODIFICATION (TMOD) NUMBER:

97 016 TITLE: AB Pumps Seal Alternate Makeup

SUMMARY

DESCRIPTION AND PURPOSE:

The Auxiliary Boller (AB) feedwater pumps are normally supplied with gland seal water from the Demineralized Water System (DM). During maintenance on the portion of the DM system that suppues the AB feed pumps, the DM line was isolated and drained. A TMOD was installed to supply Potable Water (PW) to the AB feed pump gland seals during the DM system maintenance.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was completed for this TMOD. The safety evaluation concluded that none of the systems associated with this TMOD are relied upon to mitigate the consequences of an accident. The operation of the temporary PW supply to the pump gland seats was automatic and no increased probability, consequences or malfunctions of any safety systems or accidents described in the UFSAR will occur. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

TEMPORARY MODIFICATION (TMOD) NUMBER:

97-021 TITLE:

WRGM Power During Bus 5 Outage

SUMMARY

DESCRIPTION AND PURPOSE:

This TMOD supplies an alternate non-emergency MCC source of 480 VAC power to the WRGM skid (1 RM-SKD 53) and 120 VAC power to the Control Room rad monitor system cabinet (1-RM CP 180-A). In accordance with UFSAR Section 7.5.4.4, the WRGM is normally powered from a highly reliable, diesel backed, povier source. Since the use of power from a non-emergency MCC represents a change to the facility, a safety evaluation was performed.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 safety evaluation was performed for this TMOD The determination of the safety evaluation applicability concluded that the TMOD changes the facility as described in the UFSAR (Chapters 7,.12, and 13/ Section 7.5.4.4/ Table 7,5-1).

The 480 VAC supply to the WRGM skid although not from an emergency MCC is considered to be highly reliable as there is no fuel handling accident that challenges the offsite -power system. This TMOD does' not adversely affect nuclear sefety related systems, structures, or components. The safety evaluation concluded that no unreviewod safety question was created by this TMOD.

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- TEMPORARY MODIFICATION (TMOD) NUMBER:

97 023 TITLE: Temporary Powe.' for Hot Chemistry Lab Fan AAH-FN 116 SUMi,1ARY DESCRIPTION AND PURPOSE:

This TMOD supplied temporary power to fan AAH FN 115. This fan is normally powered from a bus E-5 power supply. During the removal of bus E 5 from service ior maintenance, a source of temporary power was necessary in order to provide ventilation for exhaust hoods in the Hot Chemistry Lab.

SAFETY EVALUATION

SUMMARY

A 10CFR50.59 evaluation was completed to support this TMOD. UFSAR Figure 8.3 54 is a one line representation of electrical bus 523 that normally cupplies power to AAH FN-115. With bus 523 out of service, temporary power from bus -141 was supplied to the fan. 'ne safety evaluation concluded that this TMOD did not alter the safety function of syste.ns or components that are credited in the UFSAR.

The TMOD did not effect Erntrgency Bus E-5 that feeds bus 523 and the normal power supply was.estored prior to returning bus E-5 to service. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

TEMPORARY MODIFICATION (TMOD) NUMBER:

97-024 TITLE: Temporary Cooling Water to SI P-6A and CS-P 2A

SUMMARY

DESCRIPTION AND PURPOSE:

During the fifth refueling outage, operation of Centrifugal Charging Pump (CCP) CS-P-2A and Safety injection (SI) Pump SI-P-6A was necessary to support Emergency Core Cooling System (ECCS) flow balance testing. The normal supply of cooling water to these pumps was unavailable during the ECCS testing. This TMOD established an alternate source of cooling water for these pumps during the test.

SAFETY r. VALUATION

SUMMARY

A 10CFR50.59 evaluation was prepared to support this TMOD. During the period of time j

the ECCS flow balance test took place, the plant was in Mode 6, with the reactor core off-loaded to the Spent Fuel Pool. The CCP and SI pump were not required to be operable per Technical Specifications and they were not required to perform a safety function during the ECCS test.

The TMOD does not assure any required safety related function. The TMOD was removed and the normal cooling was realigned to the pumps before they were required to be operable or were required to perform a safety related function. The safety evaluation concluded that no unreviewed safety question was created by this TMOD.

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ENCLOSURE 3 TO NYN-97125 Listing of 10CFR50.59 Safety Evaluations Completed But Not Required Pursuant to 10CFR50.59

)

SAFETY EVALUATIONS COMPLETED RUT N_OT REQUDtED PURSUANT TO 10 CFR 5059 i

CCR 96 04 (Deletion of Valves from GL8910 Test Program)

CCR 97 02 (NUREG 0612 Polar Crane inspection Clarification)

CCR 97 04 (Requested Postponement of EDE US 64 Transformer Inspection)

MMOD 93 529 (Service Water Purnp Repair)

MMOD 93 563 (Fire Seal and Seal List Revisions)

MMOD 95 502 (Cooling Tower Fan 2 SW FN +51B Wiring Correction)

MMOD 95 513 (Replacement for Sump Pumps DF P 288A&2888)

MMOD 90 558 (RCP Cubicle Plug Removal With Polar Crane Auxiliary Hook /NUREG-0612 Response)

DCR 86 254 (Containment Equipment Hatch Bridge)

DCR 93 049 (Underground Cathodic Protection Reference Electrode Replacement)

DCR 93 052 (Turbino Generator Neutral Lead Enclosure Vontilation)

DCh 94 006 (Perimeter Security Upgrade)

DCR 94 023 (SGBR Domin Differential Pressure Switch Replacement)

DCR 95 018 (Fire Protection Storage Tanks Heating System)

DCR 95 020 (7300 Card Frame Fuse Holder Replacement)

DCR 96 001 (East Pipe Chase Heating)

DCR 96-002 (VB CP 299 Fute Alternatives)

DCR 97 018 (Containment Polar Crane Auxiliary Holst Mcdifications)

TMOD 90-030 (Use of Unit 2 Condensate Pump Motor)

TMOD 97-019 (Reactivity Compmer Data Coller:lon on MPCS)96-040 001 (Safety Evaluation for Modified Cycle 5 Fq(Z) Limits with Fixed Incore Detection System inoperable)96-002 001 (Fixed Incore Instrumentation (FICl) Thimble Insertion) 06-064-001 (Body to Bonnet Joint Torque for 1 CS FCV 121) 06-072-001 (Evaluation of EFW Flow Control Valve Surveillance in Modes 1-3)96-073 001 (Safety Evaluation for Modifind Cycle 5 Fq(Z) Limits with Fixed Incore Detec*lon System inoperable \\

96 076-001 (On Une Diesel Generator Surveliiance Evaluation)96-084 001 (Transport of Unit 2 Steam Generators over Buried Unit 1 Services and Structures) 1 of 2 1-

1 I

i SAFETY EVALUATIONS COMPLETED RULN0lMQUlMDlUASUANT To _10 cERJ0.59 l

(Continued) 96 097 001 (Service Water Traveling Water Screen Changeout) 96122-001 (Transport of Spare GSU Transformer over Buried Unit 1 Services and Structures) 97 013 001 (Cooling Tower Service Water Pumps Changeout) 07103 001 (Pressure Seal Hot Torque Verification) 97 013 002 (Evaluation of Service Environment Temperatures for MSFW Pipe Chases) 97 047 001 (PCCW Heat Exchanger Replacement NRCB 96-02 Lifting Evaluation) 97 093 001 (Evaluation for Cycle 6 RCCA Ejection Transients Without Trip)

WR97000405 (Fixed incore Instrumentation Flushing Tee Cap Tightening)

Rev. 4 Seabrook Inservice Testing Reference Manual, (Cold Shutdown Justification CSJ 8 EFW Flow Control Valve Surveillance)

Rev. 5 Seabrook Inservice Testing Reference Manual, (Cold Shutdown Justification CJC 2 PCCW Heat Exchanger Service Water Isolation) and (Deletion of Nitrogen Gas Check Valves from the IST Program) 10 CFR 50.59 Safety Evaluation Justification fu Venting the Reactor Vessel Ciosure Head Using a Temporary Exhaust Rig f

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