B16293, Annual Rept Jan-Dec 1996

Annual Rept Jan-Dec 1996

ML20136B732
Person / Time
Site:	Millstone
Issue date:	12/31/1996
From:	Bowling M NORTHEAST NUCLEAR ENERGY CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B16293, NUDOCS 9703110100
Download: ML20136B732 (101)

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Northeast ""P' F"Y "* ("a"'" *h **'ed"nt enxa85 Nuclear Energy shu.ione Nnaear eower sianon Northeet Nuclear Energy Company P.O. Box 128 Waterioni, Ur 06385-0128 (860) 447-1791 Fax (860) 444-4277 FE828m The Northeast titilitien System Docket No. 50-336 B16293 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 2 Annual Report Pursuant to the provisions of 10CFR50.59, and Section 6.9.1.4 and 6.9.1.5 of Appendix A to DPR-65, this report is submitted covering operations for the period l January 1,1996 to December 31,1996. l l

Should you have any questions regarding this report, please contact Mr. R. G. Joshi at (860)440-2080.

Very truly yours, l NORTHEAST NUCLEAR ENERGY COMPANY FOR: Martin L. wling, Jr.

Millston nit No. 2 Recovery Officer BY: .w J. A ribe Dir or, Millstone Unit No. 2 Enclosure cc: H. J. Miller, Region I Administrator D. G. Mcdonald, Jr., NRC Senior Project Manager, Millstone Unit No. 2 D. P. Beaulieu, Senior Resident inspector, Millstone Unit No. 2 ',

Dr. W. D. Travers, Director, Special Projects Office l

,L0 Director, Office of Nuclear Regulatory Research D U.S. Nuclear Regulatory Commission -

Washington, DC 20555 Attention: REIRS Project Manager I

9703110100 961231 DR ADOCK 0500 6 l

Docket No. 50-336 Millstone Nuclear Power Station Unit No. 2 _

3 Annual Report January 1,1996 through December 31,1996 r

w, Rope Ferry Rd. (Route 156). Ytterford. CT 06385 N::rtheast CeMDMg Millstone Nuclear Power Station l

Northeast Nuclear Energy Company P.o. Box 128 Waterford, CT 06385-0128 (860) 447 1791 Fax (860) 444 4277 FB28m The Northeast Utilities System Docket No. 50-336 B16293 l

U.S. Nuclear Regulatory Commission Attention: Document Control Desk l Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 2 Annual Report -

Pursuant to the provisions of 10CFR50.59, and Section 6.9.1.4 and 6.9.1.5 of Appendix A to DPR-65, this report is submitted covering operations for the period January 1,1996 to December 31,1996. ,

1 Should you have any questions regarding this report, please contact Mr. R. G. Joshi at .

l (860)440-2080.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY i FOR: Martin L. wiing, Jr.

Millston nit No. 2 Recovery Officer BY: .A A J. A~ ri'ce Dir or, Millstone Unit No. 2 Enclosure cc: H. J. Miller, Region I Administrator D. G. Mcdonald, Jr., NRC Senior Project Manager, Millstone Unit No. 2 D. P. Beaulieu, Senior Resident inspector, Millstone Unit No. 2 Dr. W. D. Travers, Director, Special Projects Office Director, Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: REIRS Project Manager 0s3422 5 REY.12-95

Docket No. 50-336 i

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Millstone Nuclear Power Station 4

! Unit No. 2 -

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l l Annual Report l

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i January 1,1996 through December 31,1996 i

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MILLSTONE UNIT 2 TABLE OF CONTENTS Section P_ age l INTR ODU CTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1.

PLANT DE SIGN CHANGES . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . 2 Plant Design Change Records (PDCR)/ Design Change Records (DCR) .... 4 Final Safety Analysis Reports Change Requests (FSARCR).... ....... .. ....... 22 j Software Implementations .......... ...... .... ....... . .... ... ........ .... .. . . . . . . . . . 26 l Stand Alone Safety Evaluations .................................. .......... .. ..... .. ....... 28 PROCEDURE CHANGES . . . . . . .. .. . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 JUMPERS-LIFTED LEADS-BYPASSES .. . ... ... ........ ....... .... ... ...... ........ 66 TESTS....................................................................................................... 81 TECHNICAL REQUIREMENTS MANUAL CHANGES .............. ................... 90 EXP ERIMENT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .93 ........

CHALLENGES TO RELIEF / SAFETY VALVES . ............. ................. ...... ... 94 STEAM GENERATOR TUBING INSERVICE INSPECTION...... .................... 95 PRIMARY COOLANT IODINE SPIKING ................. ............ ....... ........ ....... 96 REGULATORY GUIDE 1.16 REPORT FOR 1996 ............................................ 97 MP2 i

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INTRODUCTION 1

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None of the plant design changes, procedure changes, jumpers lifhd leads-bypasses, tests, technical requirements, or experiments described herein constitute, nor constituted, an unreviewed safety question per the criteria of 10CFR50.59, 1

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PLANT DESIGN CHANGES I PDCR Number Iitic l

l 2-007-85 Control Room Habitability Modification 2-003-91 Control Room Heating, Ventilation and Air Conditioning (HVAC) Radiation Monitor Modification 2-155-92 4.16 KV Bus Undervoltage/Overcurrent Coordination 2-203-92 Radiation Monitors 9799A and B Alarm Indication Modification 2 043-93 Rev.1 Se vice Water (SW) Piping Replacement Phase 5 l i

2-019-95 Main Generator Stator Leak Monitoring System (SLMS)

Installation 2-046-95 Modification to Instrument Air Compressor F-3C Forced Air Cooling 2-050-95 Modification to Instrument Air Compressor F-3C Suction Line 2-053-95 Sodium Hypochlorite (NaOCl) Delivery System Redesign 2-066-95 Control Room Air Conditioning (CRAC) Compressor Head Pressure Control System DCR Number Iitig M2-96018A Service Water (SW) Strainer Backwash Piping M2-96020A Containment Recirculation Sump Screen Enclosure Redesign M2-96052 Instrument Air Dryer T52C/D Control Panel Replacement M2-96053 B Emergency Diesel Generator (EDG) Improvements M2-96057 2-HV-203B Damper Motor Replacement MP2 2

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PLANT DESIGN CHANGES (continued)

FS ARCR Number I!tig j l

95-MP2-37 Change to Containment Isolation Valves / Penetrations,  !

Section 5 1

96-MP2-43 Wide Range Logarithmic Nuclear Instrumentation (NI)

Channel Bistables 96-MP2-44 Wide Range Logarithmic Nuclear Instrumentation System (NIS) ChannelInoperative Alarm ._

96-MP2-45 Clarify Acceptance Criteria for the Inadvertent Boron Dilution Event i Sonware Imolementation Package Iille M2-95-13783 Interim Emergency Operating Procedure (EOP) Revisions  !

to Safety Parameter Display System (SPDS)

MP-96-03906 Software Implementation Package, Balance of Plant (BOP)

Log Data Storage Stand Alone Safety Evaluation Number litJim 2346A Deficient Studs for B Emergency Diesel Generator (EDG)

MP-96-09310 Trouble Shooting Automated Work Order (AWO) for Valve 2-RW-148B Flow Determination MP2 3

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I i PDCR Number Iitle 2-007-85 Control Room Habitability Modification Descriotion of Change This change is complete. It modified the control room and control room heating, ventilation and air conditioning (HVAC) system as follows:

• Installed three (3) new low leakage isolation dampers in the system supply and exhaust ducting to provide redundancy isolating the control room. Also, it revised the control room HVAC centrols such that redundant dampers are controlled off separate electrical trains.

• Installed new radiation detectors (redundant)in the control room HVAC system intake to i provide automatic isolation of the control room on high radiation. I e Eliminated control room HVAC economizer cycle.

• Increased control room filtration unit flow from 2000 cfm to 2500 cfm and installed new high capacity HEPA filters to accommodate this increase in filtration flow.

i e Reduced control room air infiltration by weather-stripping doors and sealing all penetrations as necessary. I Reason for Channe These modifications are required in accordance with TMI Action Plant Item III.D.3.4 " Control Room Habitability as Delineated in NUREG-0737."

Safety Evaluation This change resulted in an improvement of the system's ability to minimize the dose to the control room operators following a postulated accident. The power supply to vital equipment assumed in the safety analyses was not degraded because of this modification.

All modifications and additions were seismically designed. The electrical equipment was designed such that the single failure of any of the electric cables or electric devices would affect only one of the two redundant systems. No failures were identified which would prevent emergency power from being supplied from at least one of the two power supplies.

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2 PDCR Number Title 2-003-91 Control Room Heating, Ventilation and Air Conditioning (HVAC) Radiation Monitor Modification l l

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Descriotion of Channe l This change is complete. It installed two coaxial cable connectors inside cabinet C101, one in cable ZlRM9799A/E between RIT-9799A and RE-9799A and a second connector in~ cable l Z2RM9799A/E between RIT-9799B and RE-9799B.

Reason for Channe

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The connector provides a means of simulating loss of signal from the sensor during functional testing. The test was previously performed by disconnecting a coaxial connector on an extender j

module. Eliminating use of the extender module prevents problems with the modules multipin 4

connector. Connector pins had been damaged from arcing and in some cases pushed out through l the mating connector while installing the extender module or reinstalling the radiation monitor l module.

Safety Evaluation The connector is in the cable that connects both power and signal between the remotely mounted sensor and monitor in the control room. Installation of the connector did not change how the monitor functions nor did it introduce any new failure modes. Electrically, the connector became

part of the cable between the sensor and monitor. There is no increase in the probability of l connector cable combination failure.

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l PDCR Number Title 2-155-92 4.16 KV Bus Undervoltage/Overcurrent Coordination Descriotion of Channe This change is complete. It increased the Engineered Safeguards Actuation System (ESAS) loss-of-voltage (<70%) time delay from 0.5 to 2.0 seconds.

Reason for Change The loss-of-voltage time delay setting change provided for the proper coordination of undervoltage and overcurrent protective tripping functions. By providing the proper coordination for a bus fault, the overcurrent relay trips into a lockout relay thus prohibiting the EDG from closing into a bus fault.

This change provided a greater margin to protect against inadvertent tripping of the EDG for system stability concerns.

Safety Evaluation This safety analysis evaluated a 15 second total time delay for the EDG breaker closure for loss of coolant accident / loss of normal power conditions. The setting change increased the tota!

undervoltage time delay to 12 seconds Since the 12 seconds is less than and in parallel with the 15 second EDG start time, the setting change did not adversely impact the safety analysis.

There was no affect on the probability of an accident or malfunction because no new components were added which could adversely affect the probability of detecting or initiating a loss ofnormal power. All equipment responses were in accordance with the assumption of the safety analysis.

Therefore, the margin of safety was not reduced.

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PDCR Number Title 2-203-92 Radiation Monitors 9799A and B Alarm Indication Modification Descriotion of Change This change is complete. This change revised the indicating lights on control room radiation monitors RIT-9799A and B so that Red is trip 2, Yellow is trip 1, and Green is operate. Red and Yellow were reversed in the existing installation. A wiring change was necessary to use the top light for high alarm indication.

Reason for Chance ._ i This change was made to make the alarm light color and function consistent with the other radiation monitors in the control room.

Safety Evaluation There was no change to the control room heating, ventilation, and air conditioning system. A different relay contact in the same radiation monitor was used to actuate control room recirculation. The equipment function and design basis did not change.

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PDCR Number Iills 2-043-93 Rev.1 Service Water (SW) Piping Replacement Phase 5 l 1

Descriotion of Channe This change is complete. This design change performed the following:

Replaced three turbine building closed cooling water (TBCCW) inlet and six TBCCW deteriorated outlet pipe spools with solution annealed stainless steel. A seventh outlet spool was replaced with carbon steel.

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Inspected and replaced eroded three inch to six inch Cu-Ni SW supply and return piping for the switch gear room coolers and chillers.

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Installed a connection and isolation valve in the fire main so that it could be easily connected to the emergency diesel generator (EDG) SW supply. l

. Completed repair to the Insituform pipe lining system on the A SW header in wrbine building pipe tunnel.

• Installed two SW pipe spools in auxiliary building elevation -5'6" and -25'6".

Installed four SW pipe spool: with redesigned branch connections for the EDG supply piping with vents and/or instrument connections.

• Replaced the EDG heat exchanger channel heads.  ;

e Installed new QA isolation valves and orifice plates on the SW (lube water) supply piping to the circulation pumps.

Reason for Change e Replacement of the pipe spools eliminated leak sites at inlet and outlet vent connections which had been encountered during operation. The seventh outlet spool was a carbon steel replacement to climinate an outlet vent.

• Piping in high erosion / corrosion areas was inspected and replaced to ensure that the pipe had not deteriorated and would perform as designed.

• The connection to the fire main will permit one EDG to operate at its fully rated load of 2750kw.

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5 j PDCR Number Iitle  !

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2-043-93 Rev.1 Service Water (SW) Piping Replacement Phase 5

! Reason for Channe (continued) l l . In-situ form repairs were made to eliminate a temporary repair made coming out of the i 1992 refueling outage.

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• Pipe spools were replaced to complete the replacement program and to repair a damaged l pipe spool and enhance the reliabiltiy of the system. ._

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. The use of small branch connections for the vent or instrumentation connections provided j a source for corrosion and eventualleakage as the PVC coating cannot be sufficiently applied in this confined area. Replacement of the four pipe spools enhanced the reliability I of the system.

• The EDG channel heads reached their end oflife.

. A new QA boundary was created because SW is no longer required to supply the SW pump bearings. The downstream piping from the new valves / orifice was downgraded to non-QA.

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Safety Evaluation The modifications were either a one-for-one replacement or have been reviewed and in all cases the applicable criteria for deadweight, thermal, and seismic conditions have been found to be ,

acceptable. The codes used are consistent with the original plant design requirements.

The operation of the SW and Fire Water systems from a safety related standpoint were not  ;

changed based on the fact that all analysis and design, materials, installation, and testing met or j exceeded the original design requirements and that these modifications did not affect the intended i functions of the system, nor would failure of these components be of concern. An infrequently ]

performed Evolution was performed prior to the header outage in order to maintain the margin of j safety during installation. Approved procedures were in place that provided for alternate cooling {

to the affected switchgear room prior to taking SW out of service which maintained the margin of l safety during installation. Approved seismic piping configurations for de-coupling of the out of service SW header were utilized in order to maintain the seismic qualification of the operable i

header during the facility outages.  !

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PDCR Number lith 2-019-95 Main Generator Stator Leak Monitoring System (SLMS)

Installation Description of Change This change is complete. It modified the main electrical generator stator cooling system. This change installed a GE SLMS on the main generator stator cooling water skid.

- Reason for Change The MP2 main generator had a history ofleaks in the generator stator winding which had degraded with time to the point of a leak related high potential test failure- The method used for j monitoring was cumbersome and dangerous. The existing design of the stator cooling water l system did not include a means to directly assess cooling systern integrity.

i The SLMS was installed to perform two functions -

• Monitor the generator windings for water leaks.

l e Supply oxygen to the water system in order to improve water chemistry and reduce copper erosion.

Safety Evaluation The SLMS provides no control function and cannot initiate any protective function. It is a passive system, strictly providing a monitoring function. This change minimized any increase in

probability of the loss of stator cooling by the use of properly rated fittings, check valves, and l

minimal tubing lengths. The purpose of this system Ls to acquire generator winding leak information such that a controlled shutdown could be made if serious winding leaks were identified, thus preventing an unplanned turbine generator and reactor trip from full powi:r.

l It does not affect any equipment important to safety.

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PDCR Number Ijils 2-046-95 Modification to Instrument Air Compressor F-3C Forced Air Cooling Descriotion of Change This change is complete. It provides a new, independent, non-safety related and seismic Class II l

forced air cooling system for the ins'.rument air compressor F-3C.

The new cooling system provides the cooling requirement for the compressor all year round, and reduces the ambient air temperature during summer by discharging the compressor heat rejection directly outdoors. It utilizes the compressor heat rejection to supplement the building heating in winter and provides air filtration for the cooling air intake to reduce the clogging of the heat exchanger. The system includes ductwork, a duct-mounted fan, manual dampers, and air intake l and discharge louvert. This change included installation of an inlet filter DP gauge and  !

installation of an access ladder.

Reason for Chance This new forced air cooling system eliminates a potential for compressor overheating experienced in past summer months.

Safety Evaluation Classification of this cht.nge as non-safety related and seismic Class II is consistent with the instmment air compressor F-3C classification which is also non-safety related and seismic Class II. The F-3C compressor is r.ot relied upon for emergency operation and is supplied from a non-vital power supply. This change eliminates any potential overheating of compressor F-3C. This new forced air system does not perform any safety related function and is not required to mitigate consequences of any design basis accident.

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i PDCR Number Title 2-050-95 Modification to Instrument Air Compressor F-3C Suction Line l l

Descriotion of Change This change is complete. The air intake line for instrument air compressor F-3C was rerouted to exit the west wall of the turbine building on elevation 54'6" instead of elevation 14'6".

Reason for Change This modification was necessary to ensure a clean source of plant breathing air when mnning the F-3C instrument air compressor. The existing F-3C suction line was locited in the vicinity of several chemical storage tank relief / vent lines. In addition, the suction line was located approximately 10' above the roadway west of the turbine building, creating the potential for the intake of motor vehicle exhaust.

Safety Evaluation The air intake line and compressor F-3C are not safety related and are not required to perform any function to mitigate any design basis accident. This change does not impact F-3C capacity or operation and does not introduce a new failure mode for F-3C. F-3C is not relied upon for emergency operation and is supplied by non-vital power supply.

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2-053-95 Sodium Hypochlorite (NaOCl) Delivery System Redesign Descriotion ofChange .

l This change is partia'lly complete. The service water NaOCl delivery system is being replaced i

between the NaOCl storage tanks and the service water pumps. It is being replaced with a new

Hastelloy C-22 welded system. This change includes a new NaOCl delivery system in the intake structure from the supply piping from the tanks up to the service water pump columns, and a concrete curb provided in the vicinity of the new NaOCl pumps.

Egason for Change -

The purpose of this change is to replace the NaOCl with a redesigned system. The system

, redesign is to provide more accurate and more reliable delivery of NaOCI. This is based on the operating experience of the Millstone Unit 2 NaOCl system.

i Safety Evaluation l This change represents only a change to improve performance or reliability. Operation is not affected. The changes do not revise the design bases, function or licensing bases of the system.

The NaOCl is not safety related and is not addressed in Millstone Unit 2 Technical Specifications.

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i PDCR Number Title 2-066-95 Control Room Air Conditioning (CRAC) Compressor Head Pressure Control System

] Descriotion of Channe This change is partially complete. This change backfitted the Facility 2 CRAC head pressure control system only. The CRAC system is being backfitted with a compressor head pressure control system. The system consists of a class IE safety related microprocessor based controller which will sense compressor discharge pressure from a transmitter and attempt to maintain it at a specified set point ofinitially 250 psig. Maintaining the pressure will be by controlling the speed of the condenser fan. --

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Reason for Channe Bechtel specification 7604-M531 (CRAC) originally required controls for the system operation l only when temperature was above 65 degrees outside. In 1985, the Economizer Mode was I eliminated from CRAC operation as a result of control room modifications for post TMI action.

As a result of the modifications, the air conditioning compressor would now have to operate at temperatures below 65 degrees. Since there was no compressor head pressure control for this operation range, compressor operation in the cold months has been troublesome. Numerous plant incident reports have been the subject of compressor shutdowns due to poor suction pressure, which is tied to this problem. This modification will provide continuous CRAC compressor reliability regardless of outside temperature. This change will increase CRAC system overall reliability and is part of the action plan to move CRAC system from Maintenance Rule Al status to A2 status.

Safety Evaluation Design operating parameters and logic of CRAC equipment are not changing and each of the CRAC system trains will have more reliable operation. CRAC will respond as before under accident scenarios, but with less chance of compressor system failure. The equipment is rated for all post accident environmental requirements of the CRAC room.

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DCR Number Title M2-96018A Service Water (SW) Strainer Backwash Piping Descriotion of Change This change is complete. The SW self cleaning strainer backwash piping as currently designed includes a backwash line from each strainer (L-1 A, L-1B, and L-lC) each of which ties in to a common backwash discharge header. The common header then discharges to the trough in the floor of the intake structure and out of the intake structure to the circulating water fish basket.

l This DCR modified the SW self cleaning strainer backwash piping by providing a separate <

backwash line for each strainer. The newly designed backwash lines discharge to a funnel inside the intake structure (at the south end) which, in turn, drain to a trough in the floor and out of the intake structure to the Ssh basket.

Reason for Change This change addressed the freezing issue (ACR 7654) as related to the existing backwash piping design (i.e., freezing at the discharge of the common backwash header which could potentially I render two SW headers inoperable).

Safety Evaluation This modification does not contribute to any previously analyzed accident, or its consequences and it will not contribute to any new accident outside of those already analyzed. Additionally, the margin of safety is not reduced as a result of this change. There is no possibility of a malfunction of a different type than previously evaluated as a result of this change.

The change modifies the SW strainer backwash lines and upgrades a portion of the backwash piping to QA Category 1, Seismic Class 1 to ensure backwash capability during and after a seismic event. It improves the backwash capability of the service water strainers by reducing the probability of problems associated with freezing at the point of discharge from the backwash lines.

The closure time for the air operated backwash valve was increased to reduce the probability of water hammer and potential piping damage during valve closure. The ability of the valve to close is not a safety function.

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l DCR Number Iitig M2-96020A Containment Recirculation Sump Screen Enclosure Redesign Descriotion of Change This change is complete. It resulted in the redesign, fabrication and installation of an improved QA CAT 1, seismic CAT 1 containment recirculation sump screen enclosure. The basic features and location of the sump enclosure did not change. The enclosure is structurally stronger, and constmeted mainly of stainless steel with the exception of galvanized grating. The screen mesh perforation size was decreased to address potential high pressure safety injection (HPSI) throttle

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Reason for Change Engineering evaluated the physical condition of the sump screen enclosure, reviewed the design I i

basis and considered the impact of the present screen on the containment spray (CS) and HPSI systems in regard to debris versus screen mesh size or breeches in the sump enclosure.

Engineering determined that the nonconforming conditions associated with the sump screen )

enclosure were extensive and that repairs may not guarantee an acceptable stmeture. The decision was therefore made to scrap the existing containment sump enclosure and improve sump l screen reliability.

Safety Evaluation This modification does not contribute to any previously analyzed accident, or its consequences and it will not contribute to any new accident outside of those already analyzed. Additionally, the margin of safety is not reduced as a result of this change. There is no possibility of a malfunction of a different type than previously evaluated as a result of this change.

This change did not affect any design basis accident, or its consequences. The replacement of the existing sump screen enclosure with 3/16" mesh with a new enclosure with 3/32" mesh reduces the probability of CS and HPSI system components from clogging and failing. The revised structural design enhances the enclosure integrity while maintaining key parameters such as available net positive suction head to assure adequate recirculation spray and high head safety injection pump flow.

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. M2-96052 Rev. O Instrument Air Dryer T52C/D Control Panel Replacement Description of Channe i

This change is complete. It performed the following:

i e Replaced the existing T52C/D controller (CPT52).

• Installed new repressurization valve and piping between twin desiccant towers.

j e Repowered CPT52 from VR11 Ckt. 50 instead of from compressor motor starter.

e Repowered panel CPF3C local annunciator from VR11 Ckt. 50.

i l Reason for Channe l The dryer controller upgrade was part of the instrument air system improvement effort and was i implemented to improve the overall reliability of the F3C compressor and T52C/D dryer train.

The new repressurization valve was suggested by the manufacturer as a retrofit to reduce the repressurization time of the dryer.

i The previous power supply (120VAC) for CPF3C local annunciator and CPT52 controller was fed from a single phase off the F3C compressor motor leads stepped down through a

! transformer. It was noisy and ofless than desirable quality for these microprocessor based i systems. Repowering the new controller and local annunciators from VRll provides for a more reliable source of power.

. Safety Evaluation 4

This change was made to the non-vital, backup instrument air train which is primarily used as a

! motive force for valve actuation and is not used in any reactor indication, control or protective

circuit. It has no impact on any safety related functions of any valve or component. Valves for

essential processes are designed to fail open to assure proper system alignment. Valves which

must close and be capable of opening such as the hydrogen purge valves are provided with safety

. related accumulator tanks to assure of a backup air supply to allow for valve movement. Other valves required to maintain a fixed position are designed to fail"as-is."

i lt did not affect any QA Class IE equipment or systems. There was no degradation oidryer air quality (-40*F dew point). The modifications made to the dryer controller associated with the

backup instrument air compressor / dryer train did not increase the probability of any malfunctions associated with safety related equipment.

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I DCR Number Title (continued) l M2-96052 Rev. O Instrument Air Dryer T52C/D Control Panel Replacement l l

Safety Evaluation (continued) )

l This modification reduced the potentialimpact ofr e.,ated intermittent failures. Repowering j the controller from regulated non-vital instmment p m.fv'Rll is electrically cleaner than power 1 off of the motor starter of the compressor. Any electrical shorts, opens or grounds to this new controller and it's power supply would not make the instrument air dryer fail in a way that would affect safety related systems.

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l DCR Number Iillt M2-96053 B Emergency Diesel Generator (EDG) Improvements Descriotion of Channe This change is complete. It made the following changes:

• AMOT thermostat change.

• Installation of upgraded air inlet filters and the addition of a differential pressure gage.

• Installation of Kiene (indicator) valves.

. Installation ofindependent vent lines for lube oil cooler, strainer, and fdter.

• Addition oflube oil sample valve.

• Lube oil cooler shell side drain (replacement of fittings, drain piping and drain valve with an elbow).

• Replacement of the plug on the lube oil pressure gage (PI 8757) test connection with a l l

vent isolation valve.

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• Replacement ofjacket water relief valve 2-DG-81B (PSV 8744) with a 1/2" pipe plug for use during normal operation.

• Conversion of the B EDG from a dual air start distributor system to a single air start distributor system (the A EDG currently has a single air start system).

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This change was made to support operational improvement and reliability of the Unit 2 EDGs. As a result of the failure of the MP2 B EDG, an Event Review Team (ERT) recommended various design changes to support operational improvement and reliability of the Unit 2 EDG's. Changes 1-5 were ERT recommended changes, changes 6-8 were System Engineer recommended changes, and change 9 was a manufacturer recommended change.

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I DCR Number Iitic (continued)

M2-96053 B Emergency Diesel Generator (EDG) Improvements i

1 Safety Evaluation I

All changes were enhancements to the B EDG They did not change the consequences ofloss of lube oil, EDG trip due to low lube oil pressure, or failure to start due to distributor failure. The changes were limited to a single EDG. There were no adverse impacts on the margin of safety.

Modifications improved the reliability of the EDG by :

. Lowering the normal operating temperature thereby increasing the normal lube oil pressure. .-

e Reducing the possibility of foreign material intrusion into the lube oil system. The additional pressure drop in the combustion air inlet system as a result of the new filters is well within the normal operating parameters and will be monitored with an installed filter DP gauge.

• Allowing for additional performance monitoring of the engine to aid preventive maintenance.

• Enhancing venting capability of the lube oil system to reduce air entrainment thus reducing the possibility of component failure.

• Reducing the possibility ofjacket water system leaks.

• Reducing the potential for lube oil system leakage.

• Providing a more controlled method to vent the lube oil pressure gage.

• Providing a more reliable air start distributor system.

l MP2 20

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5 DCR Number Iille M2-96057 2-HV-203B Damper Motor Replacement Descriotion of Chanad 4

This change is complete. Control room air conditioning (CRAC) F21B discharge damper 2-HV-203B motor operator was replaced with a new motor. This motor P/N M644A-E was replaced with Honeywell model M640A-1022. The new motor is a different design. It was mounted and evaluated differently, and wire cable runs were moved to accommodate the new design. All functions and circuitry remained the same. The new motor meets the same requirements as the existing motor, for seismic, electrical, and other programmatic qualifications.

Reason for Change The existing motor had failed and was no longer available from the manufacturer, Honeywell.

The plant was in a 7-day limiting condition for operation while the CRAC system was inoperable due to the broken damper operator.

Safety Evaluation All functions and circuitry of the replacement motor remained the same. Design operating I parameters and logic of CRAC equipment were not changed. The new motor meets the same .

1 requirements as the existing motor for seismic, electrical, and other programmatic qualifications.

The motor has no detrimental effects on other equipment. It is designed and qualified to meet accident environmental conditions and all safety functions.

Changing out the motor to a slightly larger model did not increase the chance of a component failure or system failure. It is expected that the new motor will result in a more reliable system.

MP2 21

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! FSARCR Number Iitig

) 95-MP2-37 Change to Containment Isolation Valves / Penetrations, Section 5 i

Descriotion of Channe -

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This Final Safety Analysis R(port (FSAR) change made changes to tables and figures in section 5 related to the containment stmeture penetration valves, pipes, and their descriptions. The affected

, items are Tables 5.2-11, 5.2-12, & 5.2-13, and Figures 5.2-27,5.2-29,5.2-32, & 5.2-38. The

! changes corrected typographical errors, made additions of components that are included in the l leakage testing progrun, corrections of other previously omitted components, and corrections to

! the figures to accurately reflect plant conditions. The accuracy of the changes was veri 6ed by

comparison with the associated P&ID drawings and valve information. ,

)

Reason for Channe '

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. This change was made because reviews showed that several FSAR section 5 tables and figures i were found to be incorrect when compared with other design documents. Section 5 in the FSAR deals with structures, and specifically with containment in section 5.2 and includes details on penetrations and associated isolation valves.

Safety E.ypluation This change is of an administrative nature and corrected minor discrepancies in various tables and figures that are used to depict the containment penetrations, isolation valves, and penetrations testing arrangements. The change was not the result of any hardware or system changes, nor did it affect test methods, boundaries, or test acceptance criteria.

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MP2 22

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FSARCR Number Iitig u l 96-MP2-43 Wide Range Logarithmic Nuclear Instrumentation (NI) ,

l Channel Bistables Description of Channe This change modified the first sentence and added two sentences in the second paragraph ofpage 7.5-9 of FSAR Section 7.5.2.4.1 Wide Range Logarithmic Channel Description. The wording addressed the fact that there are four wide range NI bistables, including a bistable which enables the control element assembly (CEA) motion inhibit power dependent insertion limit function at 10" % power and a bistable which enables the extended range mode of operation of the wide range logarithmic nuclear instrumentation channel at a very low power level. .

The change also modified the second sentence of the third paragraph of Page 7.5-9 of FSAR Section 7.5.2.4.1 to add the reactor coolant pump underspeed trip to the list of trips affected by the zero power mode bypass.

Reason for Change The FSAR, in section 7.5.2.4.1, previously described only two of the four bistable functions. The change added the description of the other two bistables. The FSAR, in section 7.5.2.4.1, previously described only two of the three reactor trips affected by the zero power mode bypass.

Safety Evaluation P

This change to the FSAR text clarified the description of the wide range nuclest instrumentation and corrected an error in the description concerning the number of bistables. The function of the I wide range nuclear instrumentation and the wide range nuclear instrumentation equipment are not affected by the change in the FSAR text.

MP2 23

J FSARCR Number Iltla 96-MP2-44 Wide Range Logarithmic Nuclear Instrumentation System (NIS) ChannelInoperative Alarm Descrintion of Change This change added two sentences after the second sentence in the second paragraph of page 7.5-9 ofFSAR Section 7.5.2.4.1, Wide Range Logarithmic Channel Description. The wording addressed the fact that the NIS Channel Inoperative alarm has been rendered inoperable by a bypass jumper (BJ), and a periodic surveillance of the wide range logarithmic channels has been implemented to compensate for the loss of the alarm.

Reason for Change The FSAR change modifies the text to reflect the plant configuration for Mode 5 and core ofiload.

Safety Evaluation The implementation of the BJ did not create an accident or condition which is not bounded by the existing design bases. This change did not introduce any credible malfunctions, and eliminated a potential failure mode affecting multiple WRNI channels. The FSAR wording changes reflected the addition of this BJ, and did not cause any malfunctions nor introduce any new failure modes intc, the WRNI channels during CORE ALTERATIONS. There was no net reduction in the margin of safety caused by the bypassjumper or the associated FSAR wording changes.

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MP2 24 m

i FSARCR Number _ Title 96 MP2-45 Clarify Acceptance Criteria for the Inadvertent Boron Dilution Event j

Descriotion of Change This change clarified statements in section 14.4.6 (Boron Dilution) of the MP2 FSAR for the CVCS malfunction. Changes were made to remove the ambiguity of the current design basis analysis for an inadvertent boron dilution.

Reason for Change

~

Section 14.4.6 (Boron Dilution) of the MP2 FSAR had statements which may lead to misinterpretation of the acceptance criteria for this event in the case of response time for the operator to manually terminate the source of dilution flow. This change clarified the FSAR write up to ensure subsequent misinterpretations will be avoided.

Safety Evaluation No malfunctions are associated with correcting misleading statements on the acceptance criteria

, of the inadvertent boron dilution analysis. The actual design basis calculation and the required plant configuration did not change. This changes did not affect the facility as described in the FSAR. It added no new components and did not modify the function of any existing component.

It did not affect the consequences of a previously evaluated accident and did not impact the actual margin to safety.

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MP2 25

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d Sonware Imolementation i

Packane Iitic M2-95-13783 Interim Emergency Operating Procedure (EOP) Revisions to Safety Parameter Display System (SPDS)

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i Descriotion of Change

! This change is complete. The SPDS software and displays were changed to provide a concise display of critical plant parameters for control room operators to aid them in rapidly and reliably determining the safety status of the plant. This change was consistent with changes to the EOPs

and the Operator Training Program. The changes were as follows

1. The vital auxiliary safety function logic was changed to include one train of service water and reactor building closed cooling water.

2. Pressurizer level is allowed to exceed 80%, where applicable, to allow for adequate l 4 subcooling. Previous success path logic required pressurizer level between 20% and 80%. l

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3. The logic for the functional recovery critical safety function for containment integrity /contamment pressure control was changed to the following success path:

e two containment air recirculation (CAR) fans / coolers And one train of containment l spray,gr l . two trains of containment spray Rtason for Change This change was made to allow the operator to rapidly evaluate the critical safety functions, and l

then to take necessary actions to restore the safety function.

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The previous SPDS logic allowed for 3 operating CAR fans / coolers to satisfy the safety function.

Three CAR fans / coolers was determit cd to be incapable of maintaining containment temperature below analysis limits fo!!owing a post ilat ,d main steam line break inside containment.

1 Safety Evaluation The SPDS represents one part of an integrated emergency response capability. The SPDS software and display changes were consistent with the original intent of the SPDS to provide a concise display ofimportant plant parameters to the reactor operator during an emergency condition. The SPDS displays and logic algorithms are now consistent with revisions to the safety function status checklists of the EOPs.

MP2 26

Software Imolementation Packane Iitig MP-96-03906 Software Implementation Package, Balance of Plant (BOP)

Log Data Storage Descriotion of Change The change provided a separate data storage Sie for BOP Log reporting. The previous BOP Log data was stored via standard archive storage software.

Reason for Channe The BOP Log software was changed to accommodate the new man machE interface for the

_ plant process computer.

Safety Evaluation The BOP Log software change did not affect any previously evaluated accidents or malfunctions, and had no impact on the margin of safety. The BOP Log continues to perform its original functions as described in the FSAR and the functional specification of automatically reporting hourly plant data.

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MP2 27

Stand Alone Safety Evaluation Number Illh 2346A Deficient Studs for B Emergency Diesel Generator (EDG)

Description of Channe This safety evaluation documented degraded fasteners on a crankcase cover on the B EDG Reason for Change "B" EDG has two vertical drive covers where:

. Two of the studs on the cover located on the control side of the EDG are missing.

• Some of the studs on the cover located on the control side and the cover opposite the control side are longer than specified in the vendor's tech manual (Fairbanks Morse Diesel Operations and Maintenance Manual #S.O.250872).

Since the B EDG was declared operable, this safety evaluation was necessary to document the degraded condition of the B EDG until repairs could be performed.

Safety Evaluation There are no adverse impacts on the margin of safety due to two missing studs on the control side based on the calculation No. 96-ENG-1516C2, Revision 0 as the remaining studs maintained their integrity. The deficient studs are longer than indicated in the vendor's tech manual and were found to be acceptable per calculation. Therefore, it is concluded that the presence of studs on the cover located on the control side and the cover opposite the control side which are longer than specified in the vendor's tech manual and the two missing studs on the control side covers do not adversely impact the margin of safety as specified in the bases of the Technical Specifications.

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1 MP2 28

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l Stand Alone Safety l

Evaluation Number lilk l

MP-96-09310 Trouble Shooting Automated Work Order (AWO) for Valve 2-RW-148B Flow Determination

. Descriotion of Chance The Spent Fuel Pool Event Review Team has determined that flow through valve 2-RW-148B, I the refuel purification pumps drain, is the most probable cause of the loss of spent fuel pool water I on 11/8/96. This safety evaluation was written to support the gathering of flow m' formation through the valve. The data gathering will be done utilizing an AWO and troubleshooting plan.

Reason for Chance

! The proposed task was to determine the range of flow that would be expected through valve 2-1 RW-148B when it is opened incrementally from full closed to a maximum of one turn open with

, the refuel purification system in service. The task utilized non intrusive Dow measuring equipment, was controlled by a troubleshooting AWO and utilized the normal flow path to the aerated waste tank. The maximum flow rate was limited to approximately 5 GPM, as measured, l

and generated less than 100 gallons of radioactive waste.

Safety Evaluation It was concluded that the proposed troubleshooting task did not increase the probability of or the consequence of an event as analyzed in the safety analysis ofMillstone Unit 2. The primary reason for this determination is the fact that the purification system plays no role in the initiation of or the mitigation of any events as defined in the safety analysis. A malfunction of a different type, an increase in the flow to radwaste, was also evaluated, and found to have no impacts.

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MP2 29 t

I PROCEDURES Procedure Numb 1I . Title I AOP2502C Rev.1 Ch.1 Loss of Vital 4.16KV Bus 24C AOP 2553 Rev. 3 Ch.1 Plant Cooldown Using Natural Circulation MP 2704B Rev. 7 Ch. 4 Installation ofReactor Cavity Seal OP2267 Rev. O Secondary Plant Cleanup Following Steam Generator (SG)

Tube Failures OP2305 Rev.17 Ch.10/ Procedure Changes to OP2305 and OP2209A to Maintain Valves OP2209A Rev. 20 Ch.1 2-RW-123 and 2-RW-124 Closed I l

OP2209A Rev. 20 Ch.7 Change in Reactor Coolant System (RCS) Draindown Required for Mid-Cycle 13 Reactor Vessel Disas .ambly OP2313C Rev.1 Procedural Changes for Powering the Hyfrogen Monitoring Containment Isolation Valves (CIAS) Post LOCA Coincident with the Loss of a Vital DC Bus OP 2341 A Rev.13 Ch.9 Fire Protection System I SP 2604L Rev. 8 Low Pressure Safety Injection (LPSI) System Alignment and Valve Operability Tests, Facility 1 SP 2604M Rev. 8 Low Pressure Safety Injection (LPSI) System Alignment and Valve Operability Tests, Facility 2 SP 2606C Rev. 8 Containment Spray (CS) System Alignment, Operability and Operational Readiness Tests, Facility 1 SP2606D Rev. O Containment Spray (CS) System Alignment, Operability and Operational Readiness Tests, Facility 2 SP 2610B Rev.11 Turbine Driven Auxiliary Feed Pump (TDAFP) Operability and Operational Readiness Tests SP 2610C Rev.10 Auxiliary Feedwater (AFW) System Lineup, Valve Operability, and Operational Readiness Tests MP2 30

PROCEDURES (continued)

Procedure Number Iills SP 2610E Rev. 7 Main Steam Isolation Valve (MSIV) Closure and Main Steam Valve Operational Readiness Testing SP 261lE Rev.4 Reactor Building Closed Cooling Water (RBCCW) Valve Operability and Operational Readiness Tests - Shutdown SP 2654D Rev. 4 Emergency Diesel Generator (EDG) Pre-Lube and Air Roll SPDS M2-95-13783 Rev. O Interim Emergency Operating Procedure (EOP) Revisions to Safety Parameters Display System (SPDS) Softwaire SPROC 96-2-1 Rev. O Manual Operation of Low Pressure Safety Injection (LPSI) Valve 2-SI-645 SPROC 96-2-1 Rev.1 Manual Operation of Low Pressure Safety Injection (LPSI) Valve 2-SI-645 SPROC 96-2-2 Rev.0 Ch.1 Batch Addition of Boric Acid to Spent Fuel Pool (SFP)

SPROC 96-2-4 Rev. O Operation of Temporary 1750KW Diesel Generator MP2 31

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i Procedure Number Iille AOP2502C Rev.1 Ch.1 Loss of Vital 4.16KV Bus 24C Descriotion of Channe 2-CHW-11 is inoperable due to maintenance. To maintain control of the separation between vital and non-vital chilled water for east DC switchgear room, Bypass Jumper (J-LL-B) #2-96-073 was implemented to gag 2-CHW-11 open. This allowed the non vital chiller to cool the room under normal conditions and required 2-CHW-145 to be closed upon LNP to bus 24C (Mode 5, 6, defueled).

Procedure OP2315D is used to maintain the room temperature below its-design limit. In this case, since 2-CHW-11 will not close automatically, a temporary procedure step was added to Procedure AOP 2502C to close manual valve 2-CHW-145 (instead of 2-CHW-11, which closes via a safety injection actuation signal / loss of normal power signal) if a loss of normal power should occur during the duration ofJ-LL-B #2-96-073. 2-CHW-145 is an isolation valve between 2-CHW-11 and the non-vital chilled water piping.

Reason for Change The temporary procedure chan8e to AOP 2502C, ensures vital chilled water cooling to the east DC switchgear room as designed following an LNP.

Safety Evaluation This change did not reduce the margin of safety. The consequences associated with the identified malfunctions have no impact on the plant's protective barriers and safety limits (specifically, fuel ,

fuel clad, reactor coolant system pressure boundary, containment and distance / dose to the public),

The change allows the plant to operate with the non-vital chiller supplying all loads (as normally) and provides steps to ensure the vital system is operated as intended. Containment integrity is not challenged. The chilled water system for the east DC switchgear room would not be operable if 2-CHW-145 was not closed to isolate the vital from non vital chiller piping sections during accident conditions. However, the existing actions in the Procedure OP2315D would maintain the east DC switchgear room below its temperature design limit. The additional temporary procedural change to AOP 2502C was made to have manual valve 2-CHW-145 (instead of 2-CHW-11, which closes via a SIAS/LNP signal), closed if an LNP should occur during the duration of this bypassjumper.

MP2 32

Procedure Number .Titic AOP 2553 Rev. 3 Ch.1 Plant Cooldown Using Natural Circulation Descriotion of Change This procedure addressed necessary actions should the power operated relief valve (PORV) block valve be closed during cooldown. The block valves are opened for 7 seconds and then closed at pressurize temperatures of 525,400, and 275 degrees F to prevent thermal binding, l

Reason for Channe This procedure was changed to address necessary actions should the PORY block valve be closed during cooldown as required by DE-2-95-580 and TS-2-95-487, 4 Safety Evaluation l

The events involved with the opening of the power operated relief valve (PORV) block valves during a cooldown were reviewed and found to have no impact on the margin of safety.

This evolution would at most trigger an inadvertent opening of a PORV, which is analyzed and not limiting in this mode. The insdvertent opening of the PORV is a full power event. The loss ofinventory is well within the capability of the high pressure safety injection and charging system.

Although the repeated opening of the PORV block valve would cause additional stress on the PORV and could result in additional leakage or failure, the reactor was shutdown and was being depressurized at the time the valves were opened.

The stroking did not increase the probability of a malfunction of any equipment important to safety in that the only valves that can be impacted were the PORV blocking valves, which are not credited in the mitigation of any event, and the PORV, which is not credited for any loss during the cooldown transient. Any possibility of additionalinventory loss during the cooldown transient due to a stmek open PORV is mitigated by the fact that the PORV block valves would only be opened enough to assure that they were off their seat.

MP2 33

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Procedure Number Tills MP 2704B Rev. 7 Ch. 4 Installation ofReactor Cavity Seal i Description ofChanne This safety evaluation specifically addresses the maintenance check of the cavity seal stud torques after flood up.

This one time procedure change added the following:

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• Instruction in step 4.6 to determine set of cavity seal O-ring and adjust fasteners to allow for this.

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• Drawing correction of bolt locations in Attachment 1.

I The procedure change consists of measuring the torque values for each individual hold down stud immediately after flood up and then periodically restoring this measured value. A final torque check will be performed immediately prior to drain down.

Reason for Channe This one-time change was made to help reduce leakage following drain-down and to correct a drawing.

To reduce the possibility ofleakage during an unanticipated drain down, re-establishing the torque values at a frequency of no more than once a month during the flooded condition, in addition to just p rior to drain down is also being planned Such additional torquing of the studs will periodically re-establish the load condition of the cavity seal assembly to that first reached upon flood up.

Safety Evaluation Periodic implementation of the torque checking procedure during the flooded condition will not result in higher design loads or stresses on any of the cavity seal components than previously considered. This procedure change does not affect any previously evaluated accidents or malfunctions and will not introduce any new accidents or malfunctions. It will not increase the risk to public health and safety.

MP2 34

Procedure Number Iltla OP2267 Rev. O Secondary Plant Cleanup Following Steam Generator (SG)

Tube Failures j

i Descriotion of Channe This operating procedure provides instructions for cleanup of the secondary plant following a SG j

tube failure. The procedure provides a matrix, listing all major secondary components with the '

" following fields ofinformation: 1) component description, 2) component identification, 3) approximate cleanup volumes, 4) sample points, 5) cleanup method, 6) cleanup point (attachment point for draining), 7) column for logging pre-cleanup activity levels, and

8) column for logging post-cleanup activity levels. The matrix is used to: 1) establish the scope of the cleanup effort, 2) maintain an ongoing status of secondary cleanut i progress, and 3) give a snapshot overview of the secondary plant after cleanup is complete.

The instruction sections of the procedure provide step-by-step guidance for aligning components l

- and sections of the secondary plant to facilitate cleanup by various methods. The primary methods involve the use of the condensate system on long recycle, draining systems to the pit sumps for pumping to the condensate polishing facility and discharge to Long Island Sound (LIS),

draining the hotwell to LIS, operation of the condensate system on short recycle, draining the SGs to radwaste and miscellaneous draining of tanks and heat exchanger shells to the hotwell for i

cleanup on recycle. The procedure provides some options for the cleanup methods used based on the levels of activity reached during the tube failure.

Reason for Change This procedure satisfies requirements specified in Significant Operating Experience Report (SOER) 93-1 to develop contingency plans for control and processing oflarge volumes of contaminated water that may be generated in the secondary systems as a result of a steam generator tube rupture. Therefore, this procedure provides guidance for determining the extent o activity in the secondary plant, removing secondary plant activity and providing a snapshot of the secondary plant status for startup evaluation after a SG tube failure.

MP2 35

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l Pfocedure Number Tjlig (continued)

OP2267 Rev. O Secondary Plant Cleanup Following Steam Generator (SG)

Tube Failures l

Safety Evaluation '

This procedure does not degrade the reliability of any safety system, increase the challenge to any safety system, or introduce unwanted or unreviewed system interactions. The evolution of operating condensate and feedwater systems on recycle and draining secondary component for activity cleanup is performed under Health Physics oversight. Additionally, no equipment will be j operated outside ofits normal design parameters or in a manner that would cause it to fail to  !

perform its required function. Operations of secondary systems per this procedure is inline with  !

current operating procedures.

Other than system sampling, the use of this procedure is performed in Modes 5, 6 and defueled when plant conditions are static. All the functions performed per this procedure were previously used processes with the exception of treating the secondary system as potentially contaminated and handling secondary water as a radiological hazard. Use of this procedure will not cause a ,

change to any system interface in a way that increases the likelihood of an accident. Additionally,  !

there are no design requirements identified in the applicable cleanup process. Appropriate Health  :

Physics and Chemistry oversight is provided for in the form of prerequisites and action steps. l This procedure does not adversely impact any previously evaluated accidents or malfunctions of equipment important to safety, does not create a new unanalyzed accident or malfunction, and i does not reduce the margin ofsafety. l i

MP2 36

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l Procedure Number Iilla l OP2305 Rev.17 Chg.10/ Procedure Changes to OP2305 and OP2209A to Maintain Valves l OP2209A Rev. 20 Chg.1 2-RW-123 and 2-RW-124 Closed l l

Descriotion of Channe These procedures were changed to maintain valves 2-RW-123 and 2-RW-124 (Refuel Water l

) Purification System Suction from Refuel Pool) in the closed position during the mid-Cycle 13 ,

i l core offload and reload activities whenever the reactor vessel head is detensioned and reactor cavity water level is above the top of the reactor vessel flange. A temporary filter system may be i

used to maintain visual clarity and reduce radiation levels in the refuel pool during the core '

- ofiload and reload operations in the mid-Cycle 13 outage. _

L Reason for Channe Adverse Condition Report (ACR) number M2-96-0136 identified that approximately 31 feet of the 4"-HCC-2 refuel water purification system suction piping is not seismically supported.

4 Closing these valves during periods when the reactor vessel head is dctensioned and reactor cavity ,

water level is above the top of the reactor vessel flange will ensure that, following a scismic event,  !

the non-seismic piping will not create a leakage path from the refueling pool.

A postulated break in the refueling purification suction line during refueling operations could create a condition which rapidly decreases the water level in the refueling pool and spent fuel pool (if the fuel transfer tube is open). This could cret.te high radiation conditions within the containment due to exposure of the UGS and/or irradiated fuel assemblies, and spent fuel pool area.

l Safety Evaluation

( l The procedure changes to maintain valves 2-RW-123 and 2-RW-124 closed during periods of l time when the reactor vessel head is detensioned and reactor cavity water level is above the top of the reactor vessel flange will not create the possibility of a new type of malfunction, since this is simply not utilizing the installed purification system. It will have no effect on the margin of safety as defined in the basis for any Technical Specification requirement. It will have no impact on any previously evaluated accidents and malfunctions of equipment important to safety. It will not create an accident or malfunction of a different type than previously analyzed. The closure of valves 2-RW-123 and 2-RW-124 during periods of time when the reactor vessel head is l

detensioned and reactor cavity water level is above the top of the reactor vessel flange will eliminate a potential leakage path following a postulated seismic event.

4 MP2 37

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Procedure Number Iltic (continued)

OP2305 Rev.17 Chg.10/ Procedure Changes to OP2305 and OP2209A to Maintain Valves OP2209A Rev. 20 Chg.1 2 RW-123 and 2-RW-124 Closed l

Safety Evaluation (continued)

A temporary filter may be used in the refuel pool to maintain visual cladty and to reduce radiation levels.. The filter unit takes a suction from and discharges to the South saddle area. The installation of this filter does not affect the operation of any safety related equipment, and does not interfere with the operation of the fuel handling equipment. The use of a temporary filter in  ;

the refuel pool has been previously evaluated and does not create the possibility of a new type of I malfunction, does not create the possibility of an accident of a different type than previously evaluated and has no impact on the margin of safety.

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MP2 38

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Procedure Number Tith

! OP2209A Rev. 20 Ch.7 Change in Reactor Coolant System (RCS) Draindown Required for Mid-Cycle 13 Reactor Vessel Disassembly

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Description of Channe 1

Based upon the measured boroo concentrations from sample points throughout the RCS, it was i not necessary to drain the RCS to approximately 3 feet below the reactor vessel flange prior to detensioning the reactor vesse'. head prior to the mid-cycle 13 core offload activities.

Reason for Channe i

The change credited the borution which was performed on July 4-5,1996 as meeting the intent of License Amendment No. 201, and drained the RCS only as necessary to detension and remove the

reactor vessel head (approximately I foot below the reactor vessel flange).

The original intent of draining the RCS to approximately 3 feet below the reactor vessel flange was to minimize the volume of water in the RCS to improve the mixing, while remaining above

} the reactor vessel level corresponding to " Reduced Inventory Operations," The RCS boration performed on July 4 5,1996 and subsequent sampling have shown that the RCS boron i

concentration has actually mixed to a greater degree than assumed in the calculations provided to

the USNRC.

• Safety Evaluation i

The boration of the RCS was performed in a manner different than presented to the USNRC. The l measured boron concentration values from various points around the RCS indicated a greater amount of mixing and a significantly smaller volume of " stagnant" water volume than was assumed in the calculation provided the USNRC. The calculation was revised using the j l information from the boron concentration measurements and assuming the RCS to be drained to 4

i i approximately 1 foot below the reactor vessel flange. The revised boron conemtration requirement is significantly less than the value currently required by Technical Specification 3.9.1 l j

(Amendment No. 201). The Bases of Technical Specification 3.9.1 currently requires the refueling boron concentration to be greater than 1950 ppm, while the revised calculation indicates that the boron concentration can be as low as 1860 ppm and still maintain the original intent of l Amendment No. 201.

I It is concluded that the boration performed on July 4-5,1996, and subsequent verification of boron mixing throughout the RCS is the equivalent of the " drain and fill" boration method approved by the USNRC. Therefore, it is not necessary to drain the RCS to approximately 3 feet below the top of the reactor vessel flange and refill the RCS prior to entry into Mode 6. l MP2 39

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1 l l Procedure Number IllJg OP2313C Rev.1 Procedural Changes for Powering the Hydrogen Monitoring Containment Isolation Valves (CIAS) Post LOCA Coincident with i i the Loss of a Vital DC Bus i

Description of Chann l This procedure change provides guidance for the powering of the hydrogen monitonng 4

containment isolation valves during post accident conditions coincident with the loss of a DC bus, by installing a simple electrical jumper inside COlR. l Reason for Channe ___ ,

i In the event of a loss of one DC bus, this procedure will provide guidance in the normal i operations procedure (OP 2313C) to ensure that there is power to one complete train of hydrogen a

monitoring isolation valves in order to open one flow-path following a containment isolation actuation signal (CIAS).

l l Safety Evaluating The valves discussed in this safety evaluation are listed below:

Component Power Component Power l H2 Monitor A H2 Monitor Z1 B H2 Monitor Z2 I

Inside CTMT suction valve 2-EB-88 Z1 2-EB-89 Z2 Outside CTMT suction valve 2-AC-12 Z2 2-AC-47 Z1 4 Retum Isolation Valve 2-AC-15 21 2-AC-20 Z2 l Only 2-AC-12 and 2-AC-47 are required to be re-energized from the opposite facility. The accident mitigating function is for these valves to close on a CIAS to ensure containment j mtegrity.

i l Post accident, coincident with a loss of a vital DC bus, this procedure will allow for the energizing l of one complete facility of hydrogen monitoring sampling valves by installing a simple electrical jumper inside COlR. Since one DC bus and its associated 120V vital AC power is lost, the installation of the jumper will violate the electrical separation criteria (Z2 power in an inoperable Z1 tray,21 power in an inoperable Z2 tray). Hydrogen monitoring is credited with manual initiation within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following the accident. All actions associated with this procedure change can be performed entirely within the control room and will allow the operator to use the existing (normal) hand switches that operate each valve without any confusion or change in operating characteristic of each hand switch. In the unlikely event of a loss of a vital DC bus, the MP2 40

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Procedure Number Tills

(continued) i OP2313C Rev.1 Procedural Changes for Powering the Hydrogen Monitoring

! Containment Isolation Valves (CIAS) Post LOCA Coincident with the Loss of a Vital DC Bus i

Safety Evaluation (continued) hydrogen monitoring syrtem will be able to respond as required within the credited 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as a l

result of this procedure change. The containment isolation function of these valves is not affected

! by this procedure change and this change will not affect the consequences of any previously I

evaluated accident which requires containment isolation.

This procedure change has no affect on any systems that could create or initiate a LOCA and cannot increase the probability of occurrence of any previously evaluated accident or previously l

evaluated malfunction of equipment important to safety. The CIAS function of these sample valves will not be altered by this change. Upon receipt of a CIAS these valves will respond as originally designed.

I The chance of an inadvertent short-circuit or grounding of the available 125-VDC bus when

' installing the jumpers is prevented since the procedure change calls for the removal of the Facility 1 and Facility 2 fuseblocks prior to installing the jumper. Once the jumper is installed, only the operable power supply fuse will be reinstalled to prevent a cross-tie of facilities. The dead bus fuse is not reinstalled followingjumper installation and the operable bus fuse provides the same system protection. The jumper wire of 600V #14 was sized for the system serviced. Therefore, I neither a cross-tie of facilities nor a grounding of the system will occur during or following jumper installation. As a result of the procedural controls on the fuse removal and installation, this installation will not create a malfunction different than previously evaluated. The proceduralized bypass jumper has a signature block for the installer of the jumper and a signature block of verifier ofjumper installation prior to re-energizing the valve, in order to prevent the potential for

! improper installation.

Following an accident that does not initiate a CIAS (small break LOCA) coincident with a loss of a DC bus (credited single failure), the operators will be required to install the electrical jumper in order to sample containment. This action does not have to be immediately performed. In the event that the small break LOCA becomes a large break LOCA and initiates a CIAS, the operable i

facility will respond to close the operable containment isolation valve in its train. Since a single failure has previously been postulated resulting in this jumper installation, the operable inside and the opposite facility outside containment isolation valve will go closed upon receipt of the CIAS signal (due to the actuation auctioneering power for ESAS) and there will be no increase to the probability of malfunction of equipment important to safety.

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Procedure Numbel Ilth OP2341 A Rev.13 Ch.9 Fire Protection System Descriotion of Channe This procedure identifies the compensatory ventilation actions required to maintain the fire water pump structure less than or equal to 104 degrees F. Flexibility is provided by allowing the fire pump house door to be opened or a portable electric fan to be installed to cool the room.

Temperature is also monitored to determine the effectiveness of any compensatory actions taken.

This flexibility is necessary since the amount and effectiveness of compensatory ventilation actions is based on outside air temperature.

Reason For Channe Currently there are no steps detailing compensatory actions to take in the event ofloss or planned maintenance of the fire pump structure exhaust fan. Compensatory ventilation will allow for planned or unplanned maintenance to be performed on the fire pump structure exhaust fan while continuing to maintain the fire pump motor in an operable environment.

Safety Evaluation Performance of compensatory ventilation actions in the Sre water pump structure does not have an impact on the margin of safety since they perform the same function as the normal room exhaust ventilation fan and since the same administrative controls would be implemented if room temperature could not be maintained. These actions are performed to maintain component operability. Since operability is determined by the design rated motor temperature being met and since compensatory ventilation is likely to maintain this required temperature, these actions will have no adverse affect on the fire water pump motor. Ultimately, if the fire pump structure's temperature cannot be maintained below it's required temperature, the appropriate limiting condition for operation will be entered as further engineering evaluation is performed.

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i Procedure Number Iille SP 2604L Rev. 8 Low Pressure Safety Injection (LPSI) System Alignment and Valve Operability Tests, Facility 1 Descriotion of Channe l

1. Incorporated previous Plant Operations Review Committee (PORC) approved changes to Revision 7:

• Added step for check of 1/5 lights for main steam isolation ( MSI) modules and corrected typographical error.

• Modified section for testing low pressure safety injection (LPSrI) valves to reflect adjustment of 2-SI-306 stroke which limits flow and prevents pump nmout. LPSI injection valves no longer required to be throttled.

2. Added Section 4.5 to address technical requirements manual (TRM) requirement for cycling LPSI valves (Appendix "R").

3. In Section 4.4, added requirements to test the override feature of Facility 1 LPSI valves following safety injection actuation signal (SIAS) test.

4. Relocated check of 2-SI-306 from OPS Form 2604L-1 to OPS Form 2604L-2 to coincide with valve alignment checks. Also added that valve is mechanically limited to approximately 50% open (by plant design change record).

Reason for Channe l

This procedure revision incorporates tests to verify that LPSI valve, Loop 1 A (2-SI-615) and l LPSI valve, Loop 1B (2-SI-625) can be manually cycled. These valves will be manually operated one at a time. An opera *.or, in direct communication with the control room, will be stationed at the valve being cycled.

These tests will be performed in Operational Mode 5 or 6, or when the plant is defueled (all irradiated fuel removed from the reactor pressure vessel and stored in the spent fuel storage pool).

The tests meet the objectives of the surveillance requirement in the TRM for these valves, thus providing assurance that the valves will function as intended in the Appendix R shutdown analysis.

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.a Procedure Number Iills (continued)

SP 2604L Rev. 8 Low Pressure Safety Injection (LPSI) System Alignment and Valve Operability Tests, Facility Safety Evaluation This operations surveillance procedure (Revision 8) tests manual operation of the Facility 1 LPSI valves, thus providing assurance that these valves perform as intended in the Appendix R -

shutdown analysis. Based on the evaluation contained herein, the procedure revision does not adversely impact any previously evaluated accidents or malfunctions of equipment important to safety, does not create a new unanalyzed accident or malfunction, and does not reduce the margin of safety as dermed in the technical specifications, safety analysis, or NRC safety evaluation reports.

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i Procedure Number Iills SP 2604M Rev. 8 Low Pressure Safety Injection (LPSI) System Alignment and Valve Operability Tests, Facility 2 1

Descriotion of Channe l

1. Incorporated previous Plant Operations Review Committee (PORC) approved changes to Revision 7:

• Added step for check of 1/5 lights for main steam isolation (MSI) modules. l l

. Modified section for testing low pressure safety injection (LPSI) valves to reflect adjustment of 2-SI-306 stroke which limits flow and prevents pump runout. LPSI i injection valves no longer required to be throttled.

2. Added Section 4.5 to address technical requirements manual (TRM) requirement for l cycling LPSI valves (Appendix R). i

3. In Section 4.4, added requirements to test the override feature of Facility 2 LPSI valves follow'mg safety injection actuation signal (SIAS) test.

4. Relocated check of 2-SI-306 from OPS Form 2604M-1 to OPS Form 2604M-2 to coincide with valve alignment checks. Also added that valve is mechanically limited to approximately 50 percent open (by plant design change record).

Reason for Channe This procedure revision incorporates tests to verify that LPSI valve, Loop 2A (2-SI-635) and i

LPSI valve, Loop 2B (2-SI-645) can be manually cycled. These valves will be manually operated one at a time. An operator, in direct communication with the control room, will be stationed at the valve being cycled.

These tests will be performed in Operational Mode 5 or 6, or when the plant is defueled (all irradiated fuel removed from the reactor pressure vessel and stored in the spent fuel storage pool).

The tests meet the objectives of the surveillance requirement in the TRM for these valves, thus providing assurance that the valves will function as intended in the Appendix R shutdown l analysis.

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Procedure Number Iills (continued)

SP 2604M Rev. 8 Low Pressure Safety Injection (LPSI) System Alignment and Valve Operability Tests, Facility 2 Safety Evaluation This operations surveillance procedure (Revision 8) tests manual operation of the Facility 2 LPSI valves, thus providing assurance that these valves perform as intended in the Appendix R shutdown analysis. Based on the evaluation contained herein, the procedure revision does not adversely impact any previously evaluated accidents or malfunctions of equipment important to safety, does not create a new unanalyzed accident or malfunction, and does not reduce the margin of safety as defined in the technical specifications, safety analysis, or NRC safety evaluation reports. -

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, Procedure Number Iills l SP 2606C Rev,8 Containment Spray (CS) Synem Alignment, Operability and Operational Readiness Tests, Facility 1 l

Descriotion of Channe l

• Modified Section 4.1 to include timing of the valve opening, though not required by technical specifications and revised OPS Form 26%C-1 to reflect same.

• Enhanced Section 4.2 to clarify independent verification requirements for valve alignment checks.

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• Added Section 4.3 and OPS Form 2606C-3 to address technical requirements manual (TRM) requirement for manually cycling Facility 1 CS system valve (s) (Appendix R).

• When cycling CS header motor operated stop valve, eliminated the option to close l 2-CS-3 A. 2-CS-4A is the preferred choice since this would reduce the amount of head l

on 2-CS-4.1 A when opening the valve and would also reduce the amount of water pushed

! through to the containment spray header thus potentially reducing any accelerated corrosion of 2-CS-SA (check valve).

Reason for Channe This procedure revision incorporates a test to verify that the A CS header isolation valve,2-CS-4.l A, can be manually cycled. An operator, in direct communication with the control room, will be stationed at the valve being manually operated.

l This test will be performed in Operational Mode 5 or 6, or when the plant is defueled (all

irradiated fuel removed from the reactor pressure vessel and stored in spent fuel storage pool).

The test meets the objectives of the surveillance requirements in the TRM for this valve, thus l providing assurance that the valve will function as intended in the Appendix R shutdown analysis.

Safety Evaluation I This engineering surveillance procedure (Revision 8) tests manual operation of a power operated valve in the CS system, thus providing assurance that this valve performs as intended in the Appendix R shutdown analysis. Based on the evaluation contained herein, the procedure revision

does not adversely impact any previously evaluated accidents or malfunctions of equipment

, important to safety, does not create a new unanalyzed accident or malfunction, and does not (

reduce the margin of safety as defined in the technical speci6 cations, safety analysis, or NRC l i

j safety evaluation reports.

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Procedure Number .Titic SP2606D Rev.0 Containment Spray (CS) System Alignment, Operability and Operational Readiness Tests, Facility 2 l

Description of Channe l

e Modified Section 4.1 to include timing cf the valve opening, though not required by technical specifications and revised OPS Form 2606D-1 to reflect same.

• Enhanced Section 4.2 to clarify independent verification requirements for valve alignment checks.

e Added Section 4.3 to address technical requirements manual (TRM) requirement for manually cycling Facility 2 CS system valves (Appendix "R").

. When cycling CS header motor operated stop valve, eliminated the option to close 2-CS-3B. 2-CS-4B is the preferred choice as directed by the System Engineer, since this would reduce the amount of head on 2-CS-4.lB when opening valve and also reduce the amount of water pushed through to the CS header thus potentially reducing any accelerated corrosion of 2-CS-5B (check valve).

Reason for Channe This procedure revision incorporates tests to verify that the "B" CS header isolation valve,2-CS-4.lB, and the RWST header "B" isolation valve,2-CS-13.lB, can be manually cycled. An operator, in direct communication with the control room, will be stationed at the valve being manua!!y operated.

These tests will be performed in Operational Mode 5 or 6, or when the plant is defueled (all irradiated fuel removed from the reactor pressure vessel and stored in spent fuel storage pool).

The tests meet the objectives of the surveillance requirements in the TRM for these valves, thus providing assurance that the valves will function as intended in the Appendix R shutdown analysis.

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Procedure Number Iitic (continued)

SP2606D Rev.0 Containment Spray (CS) System Alignment, Operability and Operational Readiness Tests, Facility 2 Safety Evaluation This engineering surveillance procedure (Revision 8) tests manual operation of two power operated valves in the CS system, thus providing assurance that these valves perform as intended in the Appendix R shutdown analysis. Based on the evaluation contained herein, the procedure revision does not adversely impact any previously evaluated accidents or malfunctions of equipment important to safety, does not create a new unanalyzed accident or malfunction, and does not reduce the margin of safety as defined in the technical specifications, safety analysis, or NRC safety evaluation reports. Additionally, the minimum boration flowpath requirements of OP 2264, " Conduct of Outages," are met.

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l Procedure Number Illh SP 2610B Rev.11 Turbine Driven Auxiliary Feed Pump (TDAFP) Operability and Operational Readiness Tests Description of Change l l

This operations surveillance procedure (Revision 11) tests operation of the Terry turbine auxiliary feed pump steam supply valve and the auxiliary feed pump turbine govemor from fire shutdown panel C-10. This procedure added the following:

• A test to verify that control of the Terry turbine auxiliary feed pump steam supply valve (SV-4188) can be isolated from the control room and that the valve can be operated from fire shutdown panel C-10. Operators, in direct communication with the control room, will I

be stationed at panel C-10 and the valve.

• A test to verify that control of the auxiliary feed pump turbine governor can be isolated from the control room and that the governor can be operated from fire shutdown panel C-

10. Operators, in direct communication with the control room, will be stationed at panel C-10 and the turbine governor.

These tests will be performed in Operational Mode 5 or 6, or when the plant is defueled. The tests meet the objectives of the surveillance requirements in the Technical Requirements Manual (TRM) for these components, thus providing assurance that the components will function as intended in the Appendix R shutdown analysis.

It added Section 4.3 to address TRM requirements for TDAFP and associated components (Appendix "R").

Reason for Chance This procedure verifies that when the isolation switch for the auxiliary feed pump turbine governor at panel C-10 is placed in the " local" position, the governor cannot be ope ated from the control room and can be operated from panel C-10.

Safety EvalualiDD This operations surveillance procedure (Revision 11) tests operation of the Terry turbine auxiliary feed pump steam supply valve and the auxiliary feed pump turbine governor from fire shutdown panel C-10, thus providing assurance that these components perform as intended in the Appendix R shutdown analysis. This procedure revision does not adversely impact any previously l

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Procedure Number I 1113 (continued)

SP 2610B Rev.11 Turbine Driven Auxiliary Feed Pump (TDAFP) Operability and Operational Readiness Tests l

l Safety Evaluation (continued)

I

! evaluated accidents or malfunctions of equipment important to safety, does not create a new unanalyzed accident or malfunction, and does not reduce the margin of safety as defined in the technical specifications, safety analysis, or NRC safety evaluation reports. I l

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Procedure Number Iills i  :

SP 2610C Rev.10 Auxiliary Feedwater (AFW) System Lineup, Valve Operability, and i Operational Readiness Tests i

Descriotion of Channe This operations surveillance procedure (Revision 10) tests manual operation of the auxiliary feedwater header cross-tie valve and the steam generator auxiliary feed regulating valves, and incorporates a failure mode test of the CST to hotwell control valve.

An operator, in direct communication with the control room, will be stationed at the valve being manually cycled. These tests will be performed in Operational Mode 5 or 6, or when the plant is j defueled. The tests meet the objectives of the surveillance requirements in the Technical

Requirements Manual for these components, thus providing assurance that the components will function as intended in the Appendix R shutdown analysis.

Reason for Change The tests incorporated luo this procedure verified the following:

t

• The auxiliary feedwater header cross-tie valve (2-FW-44) can be manually cycled.

!

• The steam generator auxiliary feed regulating valves (2-FW-43 A,2-FW-43B) can be i manually cycled.

. The CST to hotwell control valve (2-CN-241) will fail to the closed position on loss of control power to the vtJve.

Safety Evaluation This operations surveillance procedure (Revision 10) tests manual operation of the auxiliary feedwater header cross-tie valve and the steam generator auxiliary feed regulating valves, and incorporates a failure mode test of the CST to hotwell control valve, thus providing assurance that these components perform as intended in the Appendix R shutdown analysis. This procedure revision does not adversely impact any previously evaluated accidents or malfunctions of equipment important to safety, does not create a new unanalyzed accident or malfunction, and does not reduce the margin of safety as defined in the technical speci6 cations, safety analysis, or NRC safety evaluation reports.

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Procedure Number .Titic SP 2610E Rev. 7 Main Steam Isolation Valve (MSIV) Closure and Main Steam j Valve Operational Readiness Testing l l

l Descriotion of Change This operations surveillance procedure (Revision 7) tests closure of the main steam isolation valves (MSIVs), steam generator atmospheric dump valves, and the steam generator blowdown isolation valves from the bottle-up panels, tests manual operation of tue MSIV bypass valves and steam generator atmospheric dump valves, and tests operation ofNo. 2 steam generator atmospheric dump valve from fire shutdown panel C-10. It added the following:  !

e Tests to verify that each MSIV (2-MS-64A,2-MS-64B) can be closed at either bot +1e-up panel, C-70A or C-70B. Operators, in direct communication with the control room, will be stationed at the bottle-up panels and the valve being tested. These tests will be performed in Operational Mode 3, with Tavg > 515F.

• A test to verify that No. I steam generator atmospheric dump valve (2-MS-190A) can be i closed at bottle-up panel C-70A and No. 2 steam generator atmospheric dump valve (2-MS-190B) can be closed at bottle-up panel C-70B. Operators, in direct communication with the control room, will be stationed at the bottle-up panels and at each valve. This test will be performed in Operational Mode 5 or 6, or when the plant is defueled (all irradiated fuel removed from the ieactor pressure vessel and stored in the spent fuel storage pool).

• A test to verify that No. I steam generator blowdown isolation valve (2-MS-220A) can be closed at bottle-up panel C-70A and No. 2 steam generator blowdown isolation valve (2-MS-220B) can be closed at bottle-up panel C-70B. Operators, in direct communication with the control room, will be stationed at the bottle-up panels and at each valve. This test l will be performed in Operational Mode 5 or 6, or when the plant is defueled.

e A test to verify that No.1 MSIV bypass valve (2-MS-65A) and No. 2 MSIV bypass valve l

(2-MS-65B) can be manually cycled. An operator, in direct communication with the control room, will be stationed at the valve being manually cycled.

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• A test to verify that No. I steam generator atmospheric dump valve and No. 2 steam generator atmospheric dump valve can be manually cycled. An operator, in direct I communication with the control room, will be stationed at the valve being manually l cycled.

• A test to verify that No. 2 steam generator atmospheric dump valve can be operated from l fire shutdown Panel C-10. Operators, in direct communication with the control room, will be stationed at Panel C-10 and the valve.

MP2 53 l I

-,wm i .,r _ w w w-..+ --

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Procedure Number Iills (Continued)

SP 2610E Rev. 7 Main Steam Isolation Valve (MSIV) Closure and Main Steam j Valve Operational Readiness Testing Descriotion of Channe (continued) i These tests meet the oojectives of the surveillance requirements in the Technical Requirements Manual (TRM) for these components, thus providing assurance that the components will function l c

as intended in the Appendix R shutdown analysis.

Beason for Change -

. These tests verify that when the isolation switch for the valve is placed in the " isolate" position at the bottle-up panel, the affected valve moves to the closed position.

. The test verifies that when the isolation switch for the valve is placed in the " isolate" position at the bottle-up panel, the affected valve moves to the closed position.

. The test verifies that when the isolation switch for the valve is placed in the " isolate" position at the bottle-up-panel, the affected valve moves to the closed position.

. The test verifies that No.1 MSIV bypass valve (2-MS-65A) and No. 2 MSIV bypass valve (2-MS-65B) can be manually cycled.

. The test verifies that No. I steam generator atmospheric dump valve and No. 2 steam generator atmospheric dump valve can be manually cycled.

. The test verifies that No. 2 steam generator atmospheric dump valve can be operated from fire shutdown panel C-10.

Lafety Evaluation This operations surveillance procedure (Revision 7) tests closure of the MSIVs, steam generator atmospheric dump valves, and the steam generator blowdown isolation valves from the bottle-up panels, tests manual operation of the MSIV bypass valves and steam generator atmospheric dump valves, and tests operation ofNo. 2 steam generator atmospheric dump valve from fire shutdown Panel C-10, thus providing assurance that these components perform as intended in the Appendix R shutdown analysis. The procedure revision does not adversely impact any previously evaluated accidents or malfunctions of equipment important to safety, does not create a new unanalyzed accident or malfunction, and does not reduce the margin of safety as defined in the technical specifications, safety analysis, or NRC safety evaluation reports.

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Procedure Number- Iitic SP 2611E Rev 4 Reactor Building Closed Cooling Water (RBCCW) Valve l Operability and Operational Readiness Tests - Shutdown 1

I Descriotion of Change This operations surveillance procedure (Revision 4) tests manual operation of twenty-four air operated valves in the RBCCW system.

These tests will be performed in Operational Mode 5 or 6, or when the plant is defueled (all irradiated fuel removed from the reactor pressure vessel and stored in the spent fuel pool). These tests meet the objectives of the surveillance requirements in the Technical Requirements Manual (TRM) for these components, thus providing assurance that the componehts will function as intended in the Appendix R shutdown analysis.

This procedure incorporated the following:

e Added Section 4.2,4.3, and 4.4 to address TRM requirement for testing specific RBCCW valves (Appendix R).

• Added Operations (OPS) Form 261IE-2 to accommodate additional testing.

I e Revised OPS Form 261lE-1 to match format change in Section 4.1 to accommodate additional testing for TRM requirements.

Reason for Change 1.2.1 This procedure revision incorporates tests to verify that the following air operated valves in I the RBCCW system can be manually cycled:

• RBCCW pumps A, B, and C suction valves from RBCCW headers A and B.

• RBCCW heat exchangers A, B, and C outlet valves to RBCCW headers A and B.

l e RBCCW pump B discharge X-Tie isolation valves to RBCCW heat exchangers A and l C.

e Containment air recirculation (CAR) coolers A through D RBCCW inlet isolation valves.

i a CAR coolers A through D RBCCW emergency outlet valves.

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Procedure Number lills (continued)

SP 261IE Rev. 4 Reactor Building C!osed Cooling Water (RBCCW) Valve Operability and Operational Readiness Tests - Shutdown Eggson for Channe (continued) 1.2.2 This procedure revision also incorporates tests to verify that the following air-operated valves in the RBCCW system can be manually cycled:

. A shutdown cooling (SDC) heat exchanger RBCCW outlet valve.

. B SDC heat exchanger RBCCW outlet valve.

1.2.3 For the tests described in Sections 1.2.1 and 1.2.2, an operator, in direct communication with the control room, will be stationed at the valve being cycled during this testing.

Safety Evaluation This operations surveillance procedure (Revision 4) tests manual operation of twenty-four air operated valves in the RBCCW system, thus providing assurance that these valves perform as intended in the Appendix R shutdown analysis. This procedure revision does not adversely

)

impact any previously evaluated accidents or malfunctions of equipment important to safety, does not create a new unanalyzed accident or malfunction, and does not reduce the margin of safety as defmed in the technical speci5 cations, safety analysis, or NRC safety evaluation reports.

Consideration of the impact of testing the SDC heat exchanger RBCCW outlet valves on the

, requirements of OP 2264, " Conduct of Outages," will ensure minimum conditions for spent fuel pool heat removal are met when the plant is defueled.

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Procedure Number Iitic SP 2654D Rev. 4 Emergency Diesel Generator (EDG) Pre-Lube and Air Roll Descriotion of Channe i

t This procedure eliminated the requirement to test EDG air stan solenoids every 2 weeks. l l

Solenoids will continue to be tested on a quarterly basis as specified in SP 21194.

New sections were created to address maintaining the EDGs lubricated while eliminating the quick stans, as reconur :nded by Diesel Generator Event Review Team in TS2-96-343 (trip fuel racks and roll with air).

Performance of this procedure requires the EDG fuel racks to be tripped before rolling the engine '

with air to ensure the engine will not start. Since the engine will not stan and automatically load, Technical Specification Action Statement 3.8.1.1 (Modes 1-4) or 3.8.1.2 (Modes 5,6) must be entered while the procedure is performed. The time period that the EDG is inoperable is expected to be approximately 10-15 minutes or.ce a month. It should be noted the practice of rolling the EDG with air is currently part of the normal operating procedure for the EDG. This practice is i

needed if the EDG is pre-lubed and not started within a 15 minute time period. Contingency actions were added in case of a signal which required use of diesel.

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. eason for Channe This procedure revision performs a pre-lube and air mil of the EDG rather thar. a pre-lube and l

quick stan of the EDG. This procedure insures the tube oil system is filled between operability runs (every 2 weeks) and minimizes the time engine bearings are not lubricated in the event of an emergency stan. This method ofinsuring the EDG is lubricated replaces a previous method which required a pre-lube and quick stan of the engine every 2 weeks. The quick start method is no longer a desirable method because the chances of damaging the EDG bearings without the opportunity for healing during an extended run are increased.

Safety Evaluation This procedure performs steps that are already in accordance with an approved operations procedure and also are in accordance with EDG vendor recommendations. The redundant EDG or offsite lines will continue to provide emergency power for accident mitigation and decay heat removal. The amount of time the EDG is inoperable is minimal (.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> per 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> / month).

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l l Procedure Nuinber Iith l (continued)

, SP 2654D Rev. 4 Emergency Diesel Generator (EDG) Pre-Lube and Air Roll .

l Safety Evaluation (continued) l In the event of a failure of the pre-lube pump to shut off, the operator performing the task can take action to open the breaker (locally) to stop the pre-lube pump. Additionally, the consequences of too much prelube is negated because the engine is rolled with air to throw out any oil that may have accumulated in the upper pistons.

The procedure contains a step for an independent person to verify the fuel rack is reset. ,

Additional assurance of proper fuel rack reset is provided by the lack ofthe EDG disabled I annunciator window. l A failure to reopen the booster isolation valves would not prevent the EDG from staning (starting .

time may be longer) in the event of a loss of normal power. A failure to reset the alarm / annunciator panel is indicated by the presence of an EDG disabled alarm / annunciator window locally, and on C08 in the control room.

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l Procedure Number Tills SPDS M2 95-13783 Rev. O Interim Emergency Operating Procedure (EOP) Revisions to Safety l

l Parameters Display System (SPDS) Sonware Descriotion of Channe The SPDS soRware and displays were changed to be generally consistent with the revisions to the EOP safety function status check (SFSC) made in December 1995, The changes were made as follows:

1. The vital auxiliary safety function logic was changed to include one train of service water and reactor building closed cooling water. This logic change was made to all EOPs, with the exception of functional recovery (which already has this logic).

2. Pressurizer level will be allowed to exeed 80 percent, where applicable, to allow for ,

adequate subcooling. This logic change was made to the EOPs for loss of coolant  !

accident, steam generator tube rupture, and excess steam demand.

3. The logic for the functional recovery critical safety function for containment i i

integrity / containment pressure control (CTPC-3) was changed to the following success path:

e two containment air recirculation fans / coolers grid one train of containment spray, or e two trains ofcontainment spray Reason for Change 1

This change was made to make SPDS soRware and displays consistent with the revisions to the i EOP SFSC made in December 1995.

Safety Evaluation The SPDS is strictly a monitoring device. The change is consistent with the original intent of the SPDS to provide a concise display ofimportant plant parameters to the reactor operator during an emergency condition. It is consistent with the safety function status checklists of the EOPs.

l The SPDS sonware and display changes do not impact the margin of safety, nor do they affect the reactor coolant system pressure boundary, the integrity of the reactor containment and the fuel / cladding matrix.

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Procedure Number Iills SPROC 96-2-1 Rev. O Manual Operation ofLow Pressure Safety Injection (LPSI) Valve 2-SI-645 l Descriotion of Channe This procedure provides instructions for the installation and use of temporary, manual operating equipment for valve 2-SI-645. This equipment will be used to operate and maintain the valve in position while investigating its current degraded condition.

This procedure uses two devices to maintain control of valve position, a temporary manual actuator assembly and a stem clamp. The temporary manual actuator assembly provides a means to operate the valve, with controlled torque, without the motor operator installed. The stem clamp holds the valve stem in position when it is not being operated.

Reason for Channe .

LPSI Valve 2-SI-645 failed to close fully when operated with the Limitorque motor operator.

The valve allowed approximately 1250 gallons per minute (gpm) flow when closed with maximum torque. During operation and attempted closure, a thermal overload alarm was received and the breaker tripped open. Investigation. to date has determined that the failure is not related to any valve components external to the valve pressure boundary. Radiography results indicate no foreign material in the valve seating area or valve body.

Safety Evaluation The installation and operation of this temporary manual actuator is safe. The devices have been designed to withstand all expected loads. Valve 2-SI-645 is currently inoperable and technical

< specification limiting condition for operation (LCO) 3.4.1.3 Action "a" is currently in effect for the B loop of shutdown cooling. The installation and operation of this temporary manual actuator does not affect any other valve or otherwise degrade existing system performance. It does not affect reactor coolant system pressure boundary or containment integrity. This SPROC is a part of corrective action initiated in response to entry into LCO 3.4.1.3.

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( Procedure Number . Title SPROC 96-2-1 Rev.1 Manual Operation of Low Pressure Safety Injection (LPSI) Valve 2-SI-645 Descriotion of Change This procedure provides instructions for the installation and use of temporary, manual operating equipment for valve 2-SI-645. This equipment will be used to operate and maintain the valve in position while investigating its current degraded condition.

This procedure uses two devices to maintain control of valve position: a temporary manual actuator assembly and a stem clamp. The temporary manual actuator assembly provides a means to operate the valve, with controlled torque, without the motor operator installed. The stem  :

clamp holds the valve stem in position when it is not beina operated.

l This revision reduced shutdown cooling (SDC) flow to approximately 1050 gpm and used total ,

flow control valve 2-SI-306 to throttle flow. Maximum tc,rque value for operation of valve 2-SI-645 was changed from 168 ft-lbs to 252 ft-lbs. It added a limitation of no more than 20 closure l evolutions when torquing 2-SI-645 to a value of greater than 168 ft-lbs. It added anew attachment 2 for recording the number of times a torque value of 168 ft-lbs was exceeded during 2-SI-645 closure evolutions. Shutdown cooling flow was throttled through early in the iteration (step 4.1.5). The operator will continue to throttle after the evolution is in manual, and a signature page is provided for each evolution over 168 degrees Fahrenheit.

l Reason for Change LPSI valve 2-SI-645 failed to close fully when operated with the Limitorque motor operator. The valve allowed approximately 1250 gpm flow when closed with maximum torque. During operation and attempted closure, a thermal overload alarm was received and the breaker tripped open. Investigation to date has determined that the failure is not related to any valve components external to the valve pressure boundary. Radiography results indicated no foreign material in the valve seating area or valve body.

This revision allowed all LPSI injection valves to be fully opened by reducing SDC flow and using SDC total flow control valve 2-SI-306 to throttle flow.

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Procedure Number Iitic (continued)

SPROC 96-2-1 Rev.1 Manual Operation of Low Pressure Safety Injection (LPSI) Valve 2-SI-645 Safety Evaluation (continued)

Valve 2-SI-645 is inoperable but functional and the technical specification limiting condition for operation is in effect for the B loop of SDC. The installation and operation of this temporary manual actuator did not affect any other valve or otherwise degrade existing system performance and did not affect reactor coolant system pressure boundary or containment integrity. This SPROC is part of corrective action initiated in response to entry into LCO 3.4.1.3. j l

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l Procedure Number Iitic SPROC 96-2-2 Rev.0 Ch.1 Batch Addition ofBoric Acid to Spent Fuel Pool (SFP)

I Description of Channe This procedurc conducts multiple batch additions of high concentration borated water to the SFP.

The lower concentrated boron water is taken from the upender section of the SFP and pumped to a tank (near the new fuel inspection pit) where the concentration is increased. The higher l concentration water is then pumped back to the SFP.

Original issue provides instructions for the addition of boric acid to the SFP to raise boron concentration from the current value of 1,950 ppm, to acceptable levels (approximately 2,100 ppm), for core off-load during the 1996 mid-cycle outage. This change allowed better mixing of tank contents, added steps to recirculate tank using existing pumps and changed required batch size from 100 to 50 lbs to aid in solubility.

Reason foi Change This procedure provides guidelines for performance of the batching activities and precautions to I i

prevent introduction of debris into the SFP. This procedure also has precautions to prevent the inadvertent pump down or siphoning of the SFP Acceptance criteria is provided to deternune 1 l

when an adequate boron concentration has been reached.

Safety Evaluation The SFP is a passive component designed to contain spent fuel and is supplemented by SFP cooling. This change will not affect the ability of the SFP cooling system to perform its function.

In the unlikely event that the hose, the pumps, or the tank fails, water could be sprayed on surrounding equipment. The pumping activities are being continually monitored, and should a problem develop, monitoring personnel would promptly stop the pump. This action would reduce water loss and flooding to a minimum (70 gpm pump).

Should a spray situation occur while the pump is running, the only safety related equipment affected would be the SFP radiation monitors and a suction register of the enclosure building filtration system. Spray shields around the batching equipment, and continuous monitoring while batching will reduce the likelihood of affecting this safety related equipment.

This procedure requires continuous monitoring of all batching activities. Batching activities

prescribed herein are performed on the SFP while there is no fuel movement. These activities will j only increase the boron concentration and result in an increase in the margin of safety.

MP2 63

Proced'ure Number . Tit]s SPROC 96-2-4 Rev. O Operation of Temporary 1750KW Diesel Generator Descriotion of Channe

' This procedure provided detailed guidance for the start up and shut down of the temporary 1750KW diesel generator as well as instructions for the required electrical bus alignments The temporary diesel generator served as a backup power source to the normal station power sources. The temporary generator could have been used in the event that the Unit 2 reactor station service transformer (RSST), normal station service transformer (NSST) and A diesel generator were to fail and the Unit 1 RSST, gas turbine and diesel generator were not available from the 14H bus. 1 The temporary diesel generator was connected to the A202 breaker which normally supplies the D circulating water pump motor. Therefore, the temporary diesel generator could have supplied the 24B bus upon closure of breaker A202 and bus 24D upon closure of breaker A410.

Providing power to bus 24D would have permitted the use of the C service water pump, C l

reactor building closed cooling water (RBCCW) pump and the B low pressure safety injection LPSI pump. In the event that these pumps were not available, the procedure permitted cross tying bus 24D to bus 24E. This provided the flexibility to operate the 'B' service water pump and the B RBCCW pump.

Loss of voltage on bus 24D sends an undervoltage signal to the Z2 engineered safeguard actuation system (ESAS). At this point ESAS will be in sequence 0 (Load Shed). When in

. sequence 0, breaker A410 cannot be closed. Therefore, the undervoltage signal will prevent bus 24B and 24D from being cross tied. As a result this procedure provided guidance to bypass the undervoltage signal.

Reason for Change The procedure provided detailed guidance for the start up and shut down of the temporary 1750KW diesel generator as well as instructions for the required electrical bus alignments.

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~ Procedure Number Ii11g  !

I-(continued) l SPROC 96-2-4 Rev. O Operation of Temporary 1750KW Diesel Generator j Safety Evaluat!% (continued)

The temporary diesel generator was provided as an alternate power source to the B diesel l generator. It could have been utilized in the event that the Unit 2 RSST, NSST, and A diesel generator were to fail arid the Unit 1 RSST, gas turbine and diesel generator were not available l

from the 14H bus. There would have been no fisel movement while the temporary diesel ]

generator was supplying the 24D and/or 24E bus as the only source of power. l l

4 Use of this procedure would have helped to mitigate the consequences of-the loss of normal 4

power event. The temporary diesel generator was not connected directly to any safety equipment. j

. The bus alignments contained in this special procedure are typical and are performed in existing operating procedures. This special procedure did not result in an event which could disable or .

impair equipment required for the safety of the plant. The temporary diesel generator output breaker and existing plant breakers would have effectively isolated any postulated fault without impacting the Facility 2 safeguards electrical bus.

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MP2 65

JUMPERS-LIFTED LEADS-BYPASSES G LL-B)

J-LL-B Number Title 2-95-133 Pressurizer Safety Valve Nuts (Inlet Flange) 2-96-002 Leak Repair Encapsulation ofFO-5004 (Flange) 2-96-003 Injection Leak Repair on Hand Hole Flange On T58B 2-96-026 Low Pressure Safety Injection (LPSI) Motor Operated Valve 2-SI-645 Stem Locking Clamp 2-96-027 Disabling ofNuclear Instmmentation System (NIS) Channel Inoperative Annunciator (C04, Window A12B) 2-96-031 Low Pressure S afety Injection (LPSI) Motor Operated Valve 2-SI-645 Stem Locking Clamp 2-96-041 H2 Purge Valve Control 2-96-052 Installation Of Temporary Diesel Generator 2-96-073 Mechanical Gag on Chilled Water Supply Header Cross-tie Control Valve (CV) 2-CHW-11 2-96-078 Temporary Service Water (SW) Sodium Hypochlorite System Removed J-LL-B Number Titla 2-95-083 Install Blind Flanges at Recovered Boric Acid Storage Tank (RBAST) Pump P92 Inlet & Outlet Pipe Flanges / Remove P92 2-95-087 Seal Weld and Leak Repair ofD Main Steam Dump (MS) to Condenser 2-MS-206 2-95-092 Clamp 1njection Leak Repair of A Steam Generator Feed Pump (SGFP) Minimum Flow Recirculation Valve 2-FW-36A 2-95-095 Turbine Casings Steam Leak Stop Via Weld MP2 66

JUMPERS-LIFTED LEADS-BYPASSES G-LL-B)

(continued)

Removed J-LL-B Number Illig (continued)

2-95-114 Encapsulation of Valve 2-ES-1.1 A's Leaking 1/4 Inch Cover Plug 2-95-116 Clamp Injection Leak Repair Upstream of Drain Valve 2-MS-353 2-95-121 Clamp Injection Repair on Hand Hole Flange on T58B 2-95-152 Clamp Injection Leak Repair of 2-MS-25C MP2 67

J-LL-B Number Iit!g 2-95-133 Pressurizer Safety Valve Nuts (Inlet Flange)

Description of Channe l

This J-LL-B has been removed. It installed ASME Section m, Class 2 fastener nuts on the two I pressurizer safety valve inlet flanges in lieu of the required ASME Section m, Class 1 nuts.

Reason for Channe j It was determined during power operation that the ASME Section M, Class 2 fastener nuts were inadvertently installed on the safety valve inlet flanges during the preceding outage. The incorrect nuts did not have all the nondestructive examination required by ASME Section E, Class 1. This fastener nuts were replaced with ASME qualified material to provide full compliance.

Safety Evaluation Visual examination and magnetic particle or liquid penetrant testing was not performed on all of the nuts as required by ASME Code. However, the J-LL-B did not affect the stmetural integrity or the mechanical function of the safety valves. The installed material met all the chemical and physical requirements of the ASME Section m, Class 1 Code. The incomplete nondestructive examination represented a minimal reduction in the required material quality assurance.

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J-LL-B Number Iiti.n 2-96-002 Leak Repair Encapsulation of FO-5004 (Flange)

Description of Channe This J-LL-B has been removed. An encapsulation assembly was injected with leak sealant to stop the flow orifice steam leaks. The encapsulation assembly consists of a clamp assembly, which encapsulates the flow orifice assembly and attaches to the upstream and downstream piping. A maximum of four injections could be utilized to stop the leakage.

Reason for Change -

A steam leak had developed in two of the 1/2 inch flange plugs of flow orifice (FO-5004), the 1 A feed water heater continuous vent line.

Safety Evaluation This encapsulation is safe and can be implemented without any effect on the normal operation of flow orifice (FO-5004), piping or ray plant system.. This work was performed on a non-nuclear portion of the plant system. It did not affect any Quality plant systems and had no direct operational affect on any plant equipment. It represents no change to the "Public Risk" as compared to the original design. ,

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J LL-B Number lills 2-96-003 Injection Leak Repair on Hand Hole Flange On T58B Descriotion ofChanne This J-LL-B has been removed. This leak repair consisted ofinstalling cap nuts and/or slotted studs, one at a tLne in place of the existing flange studs and nuts. The cap nuts and/or slotted studs were utilized with the existing injection clamp rig around the hand hole flange on T58 to stop the leakage. This will be the third injection on T58B, a maximum of four injections are allowed. ._

Reason for Channe The hand hole flange on the B moisture separator reheater 1st stage reheater drain tank, T58, continued to leak. The non-QA tank service conditions are 470 PSIG and 463 degrees j

Fahrenheit. An additional injection seal repair was performed.

l Safety Evaluation This procedure had no affect on plant equipment, or the rnalfunction of safety systems. It is on a non-nuclear portion of the plant system, therefore no accidents are applicable. This procedure represents no change to the "Public Risk" as compared to the original design.

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J-LL-B Number Tills 2-96-026 Low Pressure Safety Injection (LPSI) Motor Operated Valve

. 2-SI-645 Stem Locking Clamp t

Descriotion of Chanze This J LL-B has been removed. It installed a stem clamping dee.cc on the low pressure safety injection (LPSI) injection valve 2-SI-645. The valve then remained locked in its position while the actuator was removed for troubleshooting / corrective maintenance. Wnile the clamp was installed, it was not possible to open the valve. At any time the remaining three LPSI injection valve were functional. __

Reason for Chance

- This J-LL-B allowed troubleshooting and/or corrective maintenance to be performed on LPSI 2-a SI-645 during Mode 5 operation. During this mode of operation, all four LPSIinjection valves

are throttled open to provide cooling via shutdown cooling system. It allowed installation of a stem clamping device on the LPSI injection valve 2-SI-645 to prevent upward motion of the stem and disk.

Safety Evaluation The installation of this valve stem locking clamp did not affect any other valve or otherwise degrade existing system performance. It did not affect reactor coolant system pressure boundary or containment integrity. Adequate shutdown cooling flow was maintained. Valve 2-SI-645 was

! inoperable and a technical specification limiting condition of operation was in effect. Failure of the clamp to maintain the valve in a locked position did not invalidate the assumption of the boron dilution analysis.

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j J-LL-B Number Illic 2-96-027 Disabling of Nuclear Instrumentation System (NIS) Channel Inoperative Annunciator (CO4, Window A12B)

Description of Change This J-LL-B is installed. It temporarily liAs both the signal and common lead from annunciator CO4, window A12B which removes the annunciator 125 VDC sensing voltage from the nuclear instrumentation drawers. This J-LL-B eliminated the source of power from the annunciator circuit, which will prevent the generation of electromagnetic interference (EMI) should a failure

of the NIS power supply occur.

i Reason for Change

A recent failure of the Channel C wide range power supply caused the alarm relay to cycle which 4

resulted in the cycling of the annunciator circuitry. This switching of the annunciator 125 VDC sensing voltage created EMI which was coupled into the adjacent channel.

Eafety Evaluation The implementation of the subject J-LL-B disabled the NIS channel inoperative annunciator 4

(CO4, window A12B) by lining the field cable in annunciator cabinet RC22. This eliminates a common tie between the annunciator circuitry and the nuclear instrumentation drawers. This change was reviewed to ensure the implementation of the J-LL-B will not adversely impact any plant systems or the conclusions made in the design basis accident analysis. Based on the safety i evaluation, this change will not impact any previously evaluated accidents, does not create a new I unanalyzed accident or reduce the margin of safety.

MP2 72

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l J.LL-B Number Iitic j l

2-96-031 Low Pressure Safety Injection (LPSI) Motor Operated Valve  ;

2-SI-645 Stem Locking Clamp  !

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Descriotion of Change j This J-LL-B has been removed It installed a stem clamping device on the low pressure safety injection (LPSI) injection valve 2-SI-645. The valve then remained locked in its position while the actuator was removed for troubleshooting / corrective maintenance. While the clamp was I installed, it was not possible to open the valve. At any time the remaining three LPSI injection valve were functional. .-

Reason for Chance This J-LL-B allowed troubleshooting and/or corrective maintenance to be performed on LPSI 2-SI-645 during Mode 5 operation. During this mode of operation, all four LPSI injection valves I

are throttled open to provide cooling via shutdown cooling system. It allowed installation of a stem clamping device on the LPSI injection valve 2-SI-645 to prevent upward motion of the stem l and disk.

Safety Evaluation ,

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The installation of this valve stem locking clamp did not affect any other valve or otherwise degrade existing system performance. It did not affect reactor coolant system pressure boundary or containment integrity. Adequate shutdown cooling flow was maintained. Valve 2-SI-645 was inoperable and a technical specification limiting condition of operation was in effect. Failure of the clamp to maintain the valve in a locked position did not invalidt;e the assumption of the boron dilution analysis.

i MP2 73

J-LL-B Number Title 2-96-041 H2 Purge Valve Control 1

Description of Channe This J-LL-B is installed. It installed an electricaljumper in parallel with the four hydrogen purge valve interlocks for each of the enclosure building filtration system (EBFS) heaters, X61 A and X61B. These interlocks are contacts that open whenever any one of the four hydrogen purge valves are not closed. Opening these contacts causes the EBFS heaters to be deenergized These contacts are in series to ensure that any one of the four purge valves deenergizes the heaters.

This J-LL-B is restricted to Modes 3,4, 5 and 6 only. J-LL-B tags will be placed on the outside ofboxes X61 A and X61B.

1 Reason for Channe This jumper allows the EBFS heaters X61 A and X61B to operate (supply heat) with the containment purge valves open or closed. This J-LL-B is necessary to ensure that the EBFS heaters continue to operate whenever the associated fan is operating to ensure adequate relative  ;

humidity reduction. l l

Normally any one of the four containment hydrogen purge valves opening will turn off both EBFS  !

heaters to prevent explosion when exhausting hydrogen. A commen mode failure has been fouad with the connection between the four hydrogen purge valves and the EBFS heaters. To solve the common mode failure, during Modes 3,4, 5 and 6, the EBFS heaters will be on whenever fan F25A or F25B are in service.

Safety Evaluation The installation of the J-LL-B will not alter the performance of the EBFS heaters, and the heaters will not energize until Fan F25A or F25B is operating. Therefore, the potential of heater failure or potential of a fire has not increased. Thisjumper does not affect indication for the heaters. The jumper is restricted to Modes 3,4,5 and 6 when hydrogen generation is not credited.

k MP2 74

J-LL-B Number Tith 2-96-052 Installation Of Temporary Diesel Generator Descriotion ofChanne This J-LL-B has been removed. It installed a temporary 1750KW diesel generator on the north side of the intake structure. The generator cables were routed inside the intake structure through the south side manway. The 'D' circulating pump motor cables were lifted and spliced to the generator output cables. When the temporary generator power is required the generator output can be connected to the plant electrical system via the 24B bus breaker A202 and cross tied to the 24D bus via breaker A410.

This installation was limited to Modes 5 and 6 or Defuel. ._

Reason for Channe This J-LL-B provided an additional emergency power source while the A or B diesel generator was out of service. It also provided a defense in depth to the number of emergency power supplies available.

Safety Evaluation The installation of the temporary diesel generator and required modification to breaker A202 did not reduce the margin of safety. The temporary diesel generator had four levels of protection -

the breaker inside the diesel cubicle, breaker A202 on bus 24B, breaker A410 which is the cross-tie between the 24B and 24D, and A408 which is the cross-tie breaker between 24D and 24E busses. In addition, the temporary diesel generator has sufficient capacity to power all required j loads for the existing ofoperation (Mode 5, 6 or Defueled). When in use, the temporary diesel generator would not be credited as satisfying any technical specifications limiting conditions for operation (LCO). The unit complied with all LCO action statements. 1 1

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J-LL-B Number litle 2-96-073 Mechanical Gag on Chilled Water Supply Header Cross-tie Control Valve (CV) 2-CHW-11 Descriotion of Channe This J-LL-B is installed. It installed a gagging device on 2-CHW-11 chilled water supply header cross tie control valve.

Normal operation of the chilled water system consists of one of the two non-vital chillers, X-

' 196A or X-196B, aligned to provide chilled water to both A and B DC switchgear room air conditioning units, the access control area air conditioning unit, and the cable vault recirculation unit. Upon actuation of a safety injection actuation signal (SIAS)or a lo'ss of normal power (LNP), the vital chillers, X-169A and X169B, would be started and valve 2-CHW-11 would go closed.

This J-LL-B installed a mechanical gag on 2-CHW-11 to maintain the valve in the open position, thus allowing the non-vital chillers X196A/B to supply all ofits normal loads, including the vital "DC" switchgear rooms. Previously, X169A vital "DC" swi:chgear room compressor was supplying cooling to the east DC switchgeu room.

Reason for Channe This change provided operations with some fit,xibility in' operating the chilled water system. It allowed the non-vital chillers to cool the A DC switchgear room, in addition to all their other normal loads without having the vital chiller in continuous service. The X169A compressor (A DC switchgear room vital chiller) had to run continuously to cool the A DC switchgear room as long as 2-CHW-11 was isolated. With 2-CHW-11 gagged open, X169A could be placed back into its normal stand-by condition, and placed'in service on an as needed basis.

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s J-LL-B Numbsr lil]s (continued) 2-96-073 Mechanical Gag on Chilled Water Supply Header Cross-tie Control Valve (CV) 2-CHW-11 Safety Evaluation This change did not reduce the margin of safety. The consequences associated with the identified malfunctions had nce impact on the plant's protective barriers and safety limits (specifically, fuel , '

fuel clad, reactor ce;olant system pressure boundary, containment and distance / dose to the public). '

This change allowed the plant to operate with the non-vital chiller supplying all loads (as normally) and provided steps to ensure the vital system operated as intended. Containment integrity was not challenged. The chilled water system for the east DC switchgear room would not be operable if 2-CHW-145 was not closed to isolate the vital from non-vital chiller piping sections during accident conditions. However, the existing actions in Procedure OP2315D would maintain the east DC switchgear room below its temperature design limit. An additional procedural change to AOP 2502C was made to have manual valve 2-CHW-145 (instead of 2-CHW-11, which closes via a SIAS/LNP signal) closed if an LNP should occur during the duration of this J-LL-B.

I MP2 77

J-LL-B Number Title 2-96-078 Temporary Service Water (SW) Sodium Hypochlorite System Descriotion of Channe This J-LL-B has been removed. This J-LL-B replaced two existing J-11-Bs (2-93-071 and 2 064) which implemented the temporary sodium hypochlorite system operating under inservice test (IST) 93-10, Revision 2. Revision 3 to the IST was incorporated.

The IST revision replaced the hard piped connection between the NaOCl system and domestic l

water system with hose connections to be used if flushing for maintenance was required, as well as clarified and enhanced the operating procedure portion of the IST. Additionally, existing changes to Revision 2 of the IST were incorporated.

Ep. son for Channe J-LL-B 2-93-071 had been in place for nearly three years and had become difficult to read.

Replacing the two J-LL-Bs clarified the temporary system boundaries and interaction with existing plant systems.

Safety Evaluation IST 93-10, Revision 3 was in effect under this J-LL-B. No physical changes were being made to the temporary system except that the domestic water tie-in to the system was eliminated.

Domestic water is used only for flushing purposes to support maintenance of the temporary system. A hose connection was provided if the need for flushing arose.

The possibility ofloss ofintegrity of the pipe and/or tubing of the temporary system was considered in the design. All tubing in the intake structure was sleeved by using a second containment tube around the inner, fluid canying tube for the full length of the tubing. Any potential leakage of sodium hypochlorite would be directed by the outer containment tube either j i

to the floor or to the sodium hypochlorite room. Additionally, a relief valve, located at the pump discharge, provided protection from overpressurization in the discharge line. The SW pump motors were protected from sodium hypochlorite spray by spray shields installed in the areas l immediately adjacent to the motors. The strength and quantity of sodium hypochlorite being used l in the temporary system was unchanged from the permanent plant system.

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J-LL-B Number Ijils (continued) 2-96-078 Temporary Service Water (SW) Sodium Hypochlorite System Safety Evaluation (continued)

The performance and reliability of the SW system was enhanced by this temporary system by allowing better bio-fouling contre! within the SW system. The only significant change from the permanent plant system was the location of the injection point from the intake bay to the SW pumps suction bowls. The effect of this change on materials used in the pumps shows that this would not adversely affect the system.

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i J-LL-B Number Iit!g 2-95-083 Install Blind Flanges at Recovered Boric Acid Storage Tank (RBAST) Pump P92 Inlet & Outlet Pipe Flanges / Remove P92 2-95-087 Seal Weld and Leak Repair ofD Main Steam Dump (MS) to Condenser 2-MS-206 2-95-092 Clamp Injection Leak Repair of A Steam Generator Feed Pump (SGFP) Minimum Flow Recirculation Valve 2-FW-36A 2-95-095 Turbine Casings Steam Leak Stop Via Weid l 2-95-114 Encapsulation of Valve 2-ES-1.l A's Leaking 1/4 Inch Cover Plug i

2-95-116 Clamp Injection Leak Repair Upstream of Drain Valve 2- l MS-353 2-95-121 Clamp Injection Repair on Hand Hole Flange on T58B 2-95-152 Clamp Injection Leak Repair of 2-MS-25C Descriotion of Change l

The above listed J-LL-Bs were reported in the 1995 NRC Annual Report as installed and were I removed in 1996. ,

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TESTS Test Number Title T 87-005 Minimum Flow Recirculation Measurements IST 95-008 Containment Radiation Monitors RM-8123/8262 Flow Control Control Room Air Conditioning (CRAC) System Functional IST 96-003 Test Z2 IST 96-005 Bump Test for the Temporary Diesel Generator (DG)

IST-96-007 Emergency Diesel Generator (EDG) B Lube Oil Flush IST 96-008 B Emergency Diesel Generator (EDG) Retest MP2 81

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Test Number I1131

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T 87-005 Minimum Flow Recirculation Measurements Descriotion of Change This test procedure was used to gather baseline data for flow measurement of minimum flow recirculation from safety related pumps.

Reason for Change i

This data was needed to develop acceptance criteria for increased testing required by the 1980 l Edition, Winter 1981 Addenda of the ASME Boiler and Pressure VesserCode.

Safety Evaluation The temporary flow instrument used for this test was removed after each flow measurement was j completed. The flow instrument used ultrasound for flow measurement and thus did not affect )

i the piping pressure boundary. All procedures used to operate plant equipment during the test were previously approved surveillance procedures. No changes to these procedures were necessary. This procedure did not increase the probability of occurrence or the consequences of any analyzed accident, create the potential for an accident different than that discussed in the FSAR, or reduce the safety margin of any Technical Specification.

MP2 82

s Test Number Ti11g IST 95-008 Containment Radiation Monitors RM-8123/8262 Flow Control Descriotion of Test This procedure installed temporary test equipment to provide the method for monitoring the containment radiation monitor sample flow indicating switch (FIS-8123 and FIS-8162) frequency of operation to determine if further tuning of the control system is necessary. This test was to be performed only during Modes 1,2,3, or 4. This was a special test for PDCR 2-002-94, which was not a safety evaluation related change.

Reason for Test This test was to obtain data on the containment air monitoring system flow indicating switches to  ;

monitor the operation of the radiation monitors automatic flow control function under actual plant conditions.

Safety Evaluation This test did not impact the function of the monitor to close the containment isolation dampers, nor did it impact the function of the monitor to provide reactor coolant system leak detection.

One containment radiation monitor serviced the plant while the other facility was being tested.

Collection of data did not affect any plant operating systems or change any system / component design.

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t Test Number Iills IST 96-003 Control Room Air Conditioning (CRAC) System Functional Test Z2 Descriotion of Change This test functionally verified satisfactory post-modification operation of the B CRAC system and adjustable frequency drive (AFD) microprocessor based control system associated with the air cooled condenser unit fan, MF36B.

I In order to fully functional test the integrated system response and tune the control system, the B CRAC system was operated in full temperature control of the control room environment for a ponion of the IST, During the evolution, the A CRAC system remained operable and available for control room habitability.

Reason for Change Plant design change (PDCR) 2-066-95, "CRAC Compressor Head Pressure Control System," had installed the AFD unit on Facility 2 to improve year round reliable operation of the Millstone Unit 2 CRAC system. The AFD varies the condenser unit fan motor (MF36B) speed, based on the process error (from setpoint) of the refrigerant compressor (F22B) discharge pressure. This test was needed to functionally verify the post-modification operation.

Safety Evaluation This test was performed while the B CRAC system was declared " inoperable," and the plant was opera d in accordsnce with the provisions of the technical specifications. There were no destadve elements in the test which could jeopardize availability of the entire CRAC system.

All of the required automatic actions of both CRAC trains would still occur and there was ,

sufficient procedural / alarm response guidance to ensure control room habitability in the event of a high radiation signal.

4 MP2 84

Test Number Illig (continued)

IST 96-003 Control Room Air Conditioning (CRAC) System Functional Test Z2 Safety Evaluation (continued)

The potential to generate additional electro-magnetic interference / radio frequency interference during IST performance was evaluated and was concluded to be acceptable. The impact of testing on the electrical busses was also evaluated and concluded to be not adverse. A failure of the AFD and associated B train components were also possible but would not adversely affect operation of the operable A train components.

Inadvertent system mis-operation due to software and human performance / man-machine interface concerns was not an issue due to software controls, the Northeast Utilities software validation and verification procedure and the dual / independent verification process employed during testing.

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Test Number Titig IST T96-05 Bump Test for the Temporary Diesel Generator (DG)

Descriotion of Channe The purpose of this test was to ensure the connection from the temporary DG to the Millstone Unit 2 24B bus are phase aligned properly and the output voltage for the 4160 volt bus is adjusted to obtain at least 4160 volts. To verify the phase alignment, a bump test was performed using the B turbine building closed cooling water (TBCCW) pump. The tap setting on the transformer was adjusted to obtain greater than 4160 volts on the 24B bus. Another purpose of this test was to provide an overall review of the starting and operation of the temporary DG. This test provided an actual review of the SPROC 96-2-4 (operation of the temporary DG).-

During the test the reserve station service tranaformer (RSST) supplied the 24D bus. The 24B bus and loadcenter 22B were de-energized until the temporary DG powered the bus. Loadcenter 22D was cross-tied to loadcenter 22C and motor control center l Al located in the condensate polishing facility, was be supplied from Millstone Unit 3.

The A202 breaker which normally feeds the D circulating water pump was the breaker used to backfeed the 24B bus using the temporary DG, Reason for Channe The purpose of this test was to ensure that the connection from the temporary DG to the Millstone Unit 2 24B bus are phase aligned properly and that the output voltage for the 4160 volt bus is adjusted to obtain greater than 4160 volts. Another purpose of this test was to provide an overall review of the starting and operation of the temporary DG. This test provided an actual review of the SPROC 96-2-4 (operation of the temporary DG).

Safstv Evaluation The testing of the temporary DG did not reduce the margin of safety. The temporary DG had two levels ofprotection - the breaker inside the diesel cubicle, and breaker A202 on bus 24B. In addition, the temporary DG powers the B TBCCW pump motor to verify phase rotation and the F2 station air compressor for load testing. No safety related buses or loads were utilized during this test.

MP2 86

Test Number Iille (continued)

IST T96-05 Bump Test for the Temporary Diesel Generator (DG)

Safety Evaluation (continued)

The Facility I busses,24A and 24C, were powered from the normal station service transformer (backfeed) and were not transferred to the RSST during this test. The A DG was available and the 24E bus was aligned to bus 24C which is aligned to Millstone Unit I bus 14H. The 24D bus was transferred to the RSST, which allowed the 24B bus to be isolated for the performance of this test.

The Facility 1 (Zl) Facility 2 (Z2) shutdown components were not affected by the performance of this test.

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MP2 87

Test Number Titic IST-96-007 Emergency Diesel Generator (EDG) B Lube Oil Flush Descriotion of Channe This test conducted a flush of the lube oil system on the B EDG sufficient to remove any foreign materials prior to returning the diesel generator to service. This procedure provided guidelines for performance of flush activities and precautions to prevent introduction of debris into the diesel generator. Acceptance criteria was also provided to determine when an adequate flush was completed.

Reason for Channe -

This test was performed on the B EDG while it was out of service for major overhaul after catastrophic engine failure. This IST was in preparation for returning the diesel generator to service, but was not intended to be an operability test. The sole function of this flush was to ensure that the lube oil system was clean prior to starting the diesel generator. Operability was established by existing run in and test procedures.

Safety Evaluation All of the activities associated with this flush effort were performed on the inoperable B EDG, and had no affect on diesel generator performance subsequent to its return to service. This was assured by run in testing and other surveillance tests which required the EDG to run sufficiently long to demonstrate the adequacy ofits lubrication system prior to return to service. The criteria and equipment associated with flushing were consistent with the design of the diesel generator, such that the lube oil system was sufficiently clean to rupport start up. Performance requirements of the EDGs are defined margins of safety. Given these considerations, the margin of safety defined in tech specs were not affected.

There were no possible malfunctions that might have affected the OPERABLE facility components. The equipment affected by this activity is installed solely for the purpose of accident mitigation. The activity made no permanent changes.

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MP2 88

Test Number .Iltig B Emergency Diesel Generator (EDG) Retest

. IST-96-00P Description oftest This test performed, ard documented, the testing necessary to return the B EDG to operability following the work to return it to service after its bearing failure. This test also performs the re-testing necessary to validate numerous design changes, and to obtain information to allow procedure upgrades.

Reason for Test The failure of the B EDG crankcase bearings required a major engine overhaul. During this overhaul, numerous improvement design changes were implemented, several to result in a higher lubricating oil pressure at the upper crankcase oil header during normal operation. Due to the extensive work sco, during the engine overhaul, the vendor recommended a one-time "special" engine run-in method. .ST 96-008 performed all this "special" engine run-in, as well demonstrated the capability to carry load. Additional testing was required to verify design change acceptability, and to obtain information to allow operating procedure upgrades, to lessen the chance for another engine failure.

Safety Evaluation The IST 96-008 test was safe to perform. Testing was performed when the B EDO was not -

operable and thus not required to support plant operation. The engine work, and subsequent testing, has no effect on the electrical circuitry, or the safety related busses. Thus, any failure of the B EDG would not adversely impact plant operations, since the failed generator would be separated from the rest of the plant electrical busses.

MP2 89

Technical Reauirements Manual Changes TRMCR Number Ii1[g 95-2-5 Appendix R Limiting Conditions for Operation (LCOs) and Surveillances l

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l MP2 90

TRMCR Number _ Tills 95-2-5 Appendix R Limiting Conditions for Operation (LCOs) and Surveillances Descriotion of Change This change adds a new section to the Technical Requirements Manual (TRM) . It includes the LCO and applicable Modes (1,2, & 3) for operability of Appendix R components. An attached Table Ilists the Appendix R components. The Action requirements describe compensatory methods and times for cases where components are found to be inoperable. A Bases section provides background information.

The referenced Table I is the main part of the new section. It lists components by equipment ID teamber and name. Similar items are grouped together. The table also states the applicable operability requirements for each item. The specific compensatory measures are listed for the essociated items should they be found to be inoperable during normal power operation. Finally, the table lists the surveillance requirements for each item.

Following the table are four figures that depict the locations of the 17 Appendix R areas. These are copies oflarger scale drawings from the Appendix R compliance report. The figures also show the planned routes to follow with Appendix R emergency light units. These figures can also help understand where roving fire watches must travel for the compensatory actions.

Finally, for information and comparison, six sample pages from the existing MP1 TRM Appendix R section are included.

The new surveillance requirements are imposed for components that do not have surveillances performed under some other program, and are typically required to be completed on a refuel cycle schedule, being completed prior to the mode required (Mode 3) in a stanup following a refueling outage.

Item 5.6 in the Basis section credits fire brigade members' presence.

r MP2 91

TRMCR Number Tith (continued) 95-2-5 Appendix R Limiting Conditions for Operation (LCOs) and Surveillances Reason for Channe The addition of this new section provides the guidance for Appendix R that has not existed before. It is the administrative basis for the surveillance procedures that are to be used to demonstrate operability of Appendix R components and for the compensatory actions in cases l where a component is found to be inoperable. It is itself based on the Appendix R compliance ]

report and is consistent with the concepts outlined in the MP1 Appendix R TRM section. The change has incorporated review comments from two of the original Ap;rendix R Coordinators for

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NU, and from two members of Safety Analysis. 4 Safety Evaluation The change is considered administrative in nature because no hardware changes are associated with the new requirements, and because it establishes the bases for procedures that perform surveillances on installed equipment. The procedures themselves are to be reviewed & approved separately. The new section also provides a basis for compensatory actions to be taken in case a component is found to be inoperable. One key assumption inherent in Appendix R considerations is that no other accidents (such as a LOCA) have to be assumed to occur at the same time an Appendix R Sre is assumed to develop. The new surveillance requirements are imposed for components that do not have surveillances performed under some other program, are performed while shutdown, typically on a refueling cycle, and use basic component manipulation steps.

MP2 92

4 EXPERIMENTS There were no experiments performed under the provisions of Title 10, Code ofFederal Regulations, Section 50.59 during 1996.

I MP2 93 ll g ' v

CHALLENGES TO RELIEF / SAFETY VALVES In accordance with Technical Specification 6.9.1.5, the following is a report ofchallenges to the pressurizer power operated relief valves (PORVs) or safety valves during 1996.

There were no challenges to the pressurizer PORVs or safety valves during 1996.

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mumq STE AM GENERATOR TUBING INSERVICE INSPECTION

SUMMARY

OF EVENTS - 1996 In accordance with Technical Specification 6.9.1.5, the following is a report of steam generator tubing inservice inspections performed during 1996.

There have been no tubing inservice inspections performed during 1996.

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MP2 95

PRIMARY COOLANT IODINE SPIKING In accordance with Technical specification 6.9.1.5, the foUowing is a report ofprimary coolant iodine spiking occuring during 1996.

During 1996, the specific activity of the primary coolant did not exceed the limits stated in the Technical Specifications.

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27*Jan.1997 MILLSTONE WCLEAR POWER STATION T J20 REG. CU10E 1.16 REPCRT marxp) .

Unit t 2

~ # Of Personnel > 100 W er a .~ ~ Total Man.s ec a ....

Station Utility Contract Station Utility Contract .

Work & 4 6 Function Inservice inspection 5 0 13 37 ,

0 912 2 89 0 1 Engineering staff 8 0 2303 2564 Health Physics and Lab 27 147 102 9 0 6 4 80 Maint and Construction & O 0 0 12 Operations C 0 3 Supv and office staff Refueting Operations 0 1 14 0 178 2 0 34 4 107 Engineering Staff 7 0 6 1416 Health Physics and Lab 3 28 1966 23 0 0 0 0 35 Maint and construction .6 0 0 0 Operations 0* O 1 supy and Of fice Staff Routine Oper and Surv 2 29 1076 173 1287 11 0 2092 0 28 5687 Engineerine Staff 35 662 11736 Health Physics and Lab 40 358 4156 75 0 220 0 10 10221 Maint and Construction 36 0 0 55 operatiens 0 0 3_ ............

supv and of fice staf f 271 to 339 117 Routine Plant Maint 6 2 14 8 440 0 Engineering staf f 13 0 4432 134 3766 950 M:stth Physics and Lab 32 20 0 0 0 0 417 Maint and Construction 13 0 0 8 Operations 0 0 2 supv and Office Staff 0 6 253 0 73 Special Plant Maint 2 87 0 8 Engineering Staff 10 0 3 1101 1222 10813 Mealth Physics and Lab 26 13 166 7 0 16

'2 0 3 Maint and Construction 0 ,0 4 operations 0 0 2 ............

supy and office Staf f Weste Processing 1 10 0 0 1 2 0 642 Engineering staff 0 22 338 25 16 2325 Mealth Physics and Lab 29 251 1742 41 0 332 0 9 48 Maint and construction 15 0 0 2 operations 0 0 2 Supv and Office Staff 5 69 1719 290 2722 Job Totals 28 6748 0 2791 Engineering Staf f 98 0 67 12833 5159 33286 Health Physics and Lab 206 137 1084 0 574 0 26 10808 Maint and Construction 74 0 0 81 operations 0 13 .. ..e C m. .

Supv and Office staff m. . .** . . . . . . . . . . . m . . . . : m .. .. n . e 5449 39454 mmem me m m...m 406 142

...n1259 mmu. 08.

321

...m. . . n .

nn .wm .. . . .. m.

Uni .... .

.m t 2 T otmal. ......m.nn .

Dose of persons with <100 sAem 15 NOT INCLUDE 0 Note: Report contains i.ref ficial dose on(y MP2 97

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