JAFP-94-0150, Annual Summary of Changes,Tests & Experiments for 1993

Annual Summary of Changes,Tests & Experiments for 1993

ML20064G909
Person / Time
Site:	FitzPatrick
Issue date:	12/31/1993
From:	Harry Salmon POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
JAFP-94-0150, JAFP-94-150, NUDOCS 9403170069
Download: ML20064G909 (78)

Text

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James A. Fit 2 Patrick Nuclear Power Plant

P.O Box 41 i Lycoming. New York 13093 l 315 342-3840 i

1. > NewYorkPower

4# Authority 2@ *"'e"; " "

l March 14, 1994 l

JAFP-94-0150 b l I

j- United States Nuclear Regulatory Commission l Mail Station P1-137 Washington, D.C. 20555 4 Attention: Document Control Desk i l

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 i ANNUAL

SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS ,

FOR 1993 j

Enclosure:

1) Annual Summary of JAFNPP Changes, Tests, and Experiments for 1993

j. Enclosed is a summary of the changes, tests and experiments implemented at the James A. FitzPatrick Nuclear Power Plant during l l 1993.

This report provides the Nuclear Safety Evaluation number (e.g. )

JAF-SE-93-001) followed by a brief description of the corresponding change, test, or experiment and safety evaluation summary as required by 10CFR50.59(b) (2).

4 Ver truly yours, l /#

HARRY SALMON Jr. d i I

i HPS:SRD:bnr l Enclosure l

} cc: R. Barrett D. Lindsey M. Colomb F. Edler D. Ruddy A. Znremba B. Josiger (WPO) J. Gray (WPO) w/ enc. RMS (WPO) w/ enc.

JAFP File RMS (JAF) w/ enc.

'/

94031700 9 93/231 '

PDR ADOCK.05000333 i l R PDR I

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Introduction to the 1993 Annual 10CFR50.59 Reggrt '

l 10CFR50.59 states: l (a) (1) The holder of a license. . .may (i) make changes in I the facility as described in the safety analysis report, (ii)  !

make changes to the procedures as described in the safety l analysis report, and (iii) conduct tests or experiments not described in the safety analysis report, without prior ,

commission approval, unless the proposed change, test or experiment involves a change in the technical specifications i incorporated in the license or an unreviewed safety question. l l

It also states: )

(b) ...The licensee shall also maintain records of tests and I experiments carried out pursuant to paragraph (a) of this l section. These records shall include a written safety evaluation which provides the bases for the determination that the change, test or experiment does not involve an unreviewed safety question. The licensee shall furnish to (the NRC)..., ,

annually..., a report containing a brief description of such l changes, tests, and experiments, including a summary of the safety evaluation of each.

Unless otherwise noted, each safety evaluation concluded that the subject change, test or experiment did not:

+ Increase the probability of occurrence or the consequences of an accident or malfunction of structures, systems, or I components important to safety previously identified in the FSAR;

+ Create the possibility of an accident of or malfunction of a different type than any previously evaluated in the FSAR;

+ Reduce the margin of safety as defined in the basis fce Technical Specifications; And therefore, do not involve an unreviewed safety question as defined in 10CFR50.59.

Attachacnt 1 TABLE OF CONTENTS

• JAF-SE-88-106 Rev. 0 . . . . . . . . . . . . . . . . . . Page 4

• JAF-SE-88-108 Rev. 0 . . . . . . . . . . . .. . . . . . Page 5 4 l

JAF-SE-88-188 Rev. 0 . . . . . . . . . . . . . . . . . . Page 6 ,

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• JAF-SE-89-038 Rev. 0 . . . . . . . . . . . . . . . . . . Page 7 !

• JAF-SE-89-084 Rev. 0 . . . . . . . . . . . . . . . . . . Page 8

• JAF-SE-89-119 Rev. 0 . . . . . . . . . . . . . . . . . . Page 9 JAF-SE-89-132 Rev. 2 . . . . . . . . . . . . . . . . . . Page 10 JAF-SE-90-009 Rev. 1 . . . . . . . . . . . . . . . . . . Page 11 JAF-SE-90-069 Rev. 1 . . . . . . . . . . . . . . . . . . Page 12 JAF-SE-90-084 Rev. 0 . . . . . . . . . . . . . . . . . . Page 13 JAF-SE-90-087 Rev. 1 . . . . . . . . . . . . . . . . . . Page 14 JAF-SE-91-030 Rev. 1 . . . . . . . . . . . . . . . . . . Page 15 JAF-SE-91-035 Rev. 1 . . . . . . . . . . . . . . . . . . Page 16

• JAF-SE-91-050 Rev. O . . . . . . . . . . . . . . . . . . . Page 17 JAF-SE-91-071 Rev. 0 . . . . . . . . . . . . . . . . . . Page 18

• JAF-SE-91-113 Rev. 0 . . . . . . . . . . . . . . . . . . Page 19

• JAF-SE-91-122 Rev. 0 . . . . . . . . . . . . . . . . . . Page 20 JAF-SE-92-006 Rev. 4 . . . . . . . . . . . . . . . . . . Page 21 JAF-SE-92-059 Rev. 0 . . . . . . . . . . . . . . . . . Page 22 JAF-SE-92-081 Rev. 1 . . . . . . . . . . . . . . . . . . Page 23 JAF-SE-92-094 Rev. 1 . . . . . . . . . . . . . . . . . . Page 24 JAF-SE-92-167 Rev. 0 . . . . . . . . . . . . . . . . . . Page 25

• JAF-SE-92-176 Rev. 0 . . . . . . . . . . . . . . . . . . Page 26 JAF-SE-92-185 Rev. 0 . . . . . . . . . . . . . . . . . . Page 27 JAF-SE-92-214 Rev. 0 . . . . . . . . . . . . . . . . . Page 28 JAF-SE-92-215 Rev. 0 . . . . . . . . . . . . . . . . . . Page 29 Page '

Attccha nt 1

• JAF-SE-92-219 Rev. 0 . . . . . . . . . . . . . . . . . . Page 30 JAF-3E-92-245 Rev. 0 . . . . . . . . . . . . . . . . . Page 31 JAF-SE-92-246 Rev. 0 . . . . . . . . . . . . . . . . . . Page 32 JAF-SE-93-002 Rev. 0 . . . . . . . . . . . . . . . . . Page 33 JAF-SE-93-004 Rev. 0 . . . . . . . . . . . . . . . . . Page 34 JAF-SE-93-005 Rev. 0 . . . . . . . . . . . . . . . . . . Page 35 JAF-SE-93-006 Rev. 0 . . . . . . . . . . . . . . . . . . Page 36 JAF-SE-93-007 Rev. 0 . . . . . . . . . . . . . . . . . . Page 37 JAF-SE-93-008 Rev. 1 . . . . . . . . . . . . . . . . . . Page 38 JAF-SE-93-009 Rev. 0 . . . . . . . . . . . . . . . . . . Page 39 l

JAF-SE-93-010 Rev. 0 . . . . . . . . . . . . . . . . . . Page 40 JAF-SE-93-015 Rev. 0 . . . . . . . . . . . . . . . . . . Page 41 JAF-SE-93-016 Rev. 0 . . . . . . . . . . . . . . . . Page 42 JAF-SE-93-018 Rev. 0 . . . . . . . . . . . . . . . . . Page 43 JAF-SE-93-022 Rev. 0 . . . . . . . . . . . . . . . . . . Page 44 JAF-SE-93-023 Rev. 0 . . . . . . . . . . . . . . . . . Page 45 JAF-SE-93-033 Rev. 0 . . . . . . . . . . . . . . . . . Page 46 JAF-SE-93-034 Rev. 0 . . . . . . . . . . . . . . . . . . Page 47 JAF-SE-93-036 Rev. 0 . . . . . . . . . . . . . . . . . Page 48 JAF-SE-93-041 Rev. 0 . . . . . . . . . . . . . . . . . Page 49 JAF-SE-93-045 Rev. 0 . . . . . . . . . . . . . . . . . . Page 50 JAF-SE-93-047 Rev. 0 . . . . . . . . . . . . . . . . . Page 51 JAF-SE-93-049 Rev. 0 . . . . . . . . . . . . . . . . . . Page 52 JAF-SE-93-050 Rev. 1 . . . . . . . . . . . . . . . . . Page 53

( JAF-SE-93-051 Rev. 0 . . . . . . . . . . . . . . . . . Page 54 JAF-SE-93-053 Rev. 0 . . . . . . . . . . . . . . . . . . Page 55 l JAF-SE-93-054 Rev. 0 . . . . . . . . . . . . . . . . . . Page 56 JAF-SE-93-060 Rev. 1 . . . . . . . . . . . . . . . . . . Page 57 i

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1 Attachm:nt 1 JAF-SE-93-066 Rev. 4 . . . . . . . . . . . . . . . . . . Page 58 I JAF-SE-93-067 Rev. 0 . . . . . . . . . . . . . . . . Page 59 l JAF-SE-93-072 Rev. 2 . . . . . . . . . . . . . . . . . . Page 60 JAF-SE-93-073 Rev. 0 . . . . . . . . . . . . . . . . Page 61 JAF-SE-93-074 Rev. 0 . . . . . . . . . . . . . . . . . . Page 62 JAF-SE-93-076 Rev. 0 . . . . . . . . . . . . . . . . . . Page 63 l

JAF-SE-93-077 Rev. 0 l

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JAF-SE-93-078 Rev. 0 . . . . . . . . . . . . . . . . . Page 65 '

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JAF-SE-93-085 Rev. 0 . . . . . . . . . . . . . . . . Page 66 !

l JAF-SE-93-089 Rev. 1 . . . . . . . . . . . . .. . . . Page 67 JAF-SE-93-091 Rev. 0 . . . . . . . . . . . . . . . . Page 68 )

JAF-SE-93-092 Rev. 4 . . . . . . . . . . . . . . . . Page 69 JAF-SE-93-093 Rev. 0 . . . . . . . . . . . . . . . . . Page 70 JAF-SE-93-094 Rev. 0 . . . . . . . . . . . . . . . . Page 71 i

JAF-SE-93-097 Rev. 0 . . . . . . . . . . . . . . . . Page 72 1 1

JAF-SE-93-100 Rev. 0 . . . . . . . . . . . . . . . . Page 73 JAF-SE-93-104 Rev. 0 . . . . . . . . . . . . . . . . . Page 74 JAF-SE-93-105 Rev. 0 . . . . . . . . . . . . . . . . . . Page 75 JAF-SE-93-110 Rev. 1 . . . . . . . . . . . . . . . . . Page 76

• These Nuclear Safety Evaluations were incorporated prior to 1993. Audit of records determined that these items were omitted from previous 10CFR50.59 submittals.

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Attcchmont 1 Modification F1-88-091

• JAF-SE-88-106 Rev. O PERIMETER DETECTION UPGRADE This modification contains essential safeguards information.

There is no safety question pursuant to 10CFR50.59.

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Attacha nt 1 Modification: F1-88-102 l

• JAF-8E-88-108 Rev. O PTE CAMERA UPGRADE This modification contains essential safeguards information.

There is no safety qu.$stion pursuant to 10CFR50.59.

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l Modification: M1-88-182 j JAF-SE-88-188 Rev. O A AND B REACTOR FEED PUMP ROOM WALL REP'ACEMENT 4

9 j The purpose of this modification was to change the "A" and "B" Reactor Feed Pump Room removable concrete bloc?c Val _t to carbon

steel assemblies.

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This modification provided a replacement wa]l design that ,

I eliminates the undesirable characteristics vi tne concrete block l i design, but meets all applicable design criteria. The j

] replacement walls are designed for ease of assembly and I j disassembly to eliminate the dust, contamination, radioactive j waste, and manpower expenditure associated with the removal and reinstallation of a concrete block wall.

j The replacement walls consist of onn layer of 1/4 inch thick l l carbon steel plate, ASTM A36, mounted to a structural frame j comprised of angles and structural T-sections anchored to the existing concrete. The replacement walls are lined with a layer I of acoustic panels to minimize noise transmission into the Heater Bay. The replacement walls have been designed for ease of assembly and disassembly to facilitate removal in subsequent maintenance outages.

! Implementation of this modification does not constitute an j unreviewed safety question pursuant to 10CFR50.59.

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AttachmOnt 1 Nodification: M1-89-077

• JAF-SE-89-038 Rev. O CRD DRIVE WATER PUNP 03P-16A/B SUCTION AND DISCRARGE FLANGE DESIGN CHANGE l

This modification installed new tongue and groove design suction and discharge piping flanges in place of the existing raised face flanges. This change was necessary due to the manufacturers (Worthington Pump) modification of the pump suction and discharge flanges to the tongue and groove configuration.

Per ANSI B16.5 the tongue and groove flanges are equally suited 3

to the raised face configuration for the pressure and temperature

! of the application.

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There are no unreviewed safety questions pursuant to 10CFR50.59.

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l Attcchment 1 Modifications M1-88-139  ;

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• JAF-SE-89-084 Rev. O MISCELLANEOUS UPGRADE TO THE CD HANDLING l AND REFUELING EQUIPMENT The purpose of this modification was to replace the exiating 2 unit pushbutton pendant with a water tight 3 unit pendant which i includes an emergency stop button. This pendant controls the l CW/CCW rotation of the CRD handling platform (undervessel l carousel). This modification also upgraded the reactor cavi'y c '

lights to 2000 watts supplied from new 20 amp receptacles. l This modification does not represent any changes to the respective systems' function or configuration. The modificatians upgrades certain components in response to their application, such as the wet environment under the vessel for the pushbutton I pendant and the increased wattage required of the receptacles and i lights. These replacement components will perform the same j function as the previous components. As a result of being better  ;

suited for the specific applications, the replacement pendant l station and receptacles will decrease the probability of a )

failure of these components.

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There are no unreviewed safety questions pursuant to 10CFR50.59. '

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Nodification: F1-89-106 I i \

_ *JAF-SE-89-119 Rev. O TEST WELL INSTALLATION TO SUPPORT l j BUILDING GROUND WATER INTRUSION i

! STUDY l i

The purpose of the modification is to alleviate the intrusion of

' groundwater through floor slabs at ground floor elevations in the Heater Bay and Radwaste areas of the Plant. l 1

J This modification consisted of the installation of test wells j adjacent to the east perimeter wall of the Plant to elevation a 237'. Submersible pumps are installed at each wellpoint, with j the intent of pumping ground water from rock seams located below

, the buildings foundation. The pumped water is routed to nearby existing storm water drain lines, i

.; This modification has no safety-related function, but enhances j the function of the circumferential plant drain which is to prevent a buildup of groundwater pressure on the structural slabs i at ground floor elevations inside the Plant.

J i There are no unreviewed safety questions pursuant to 10CFR50.59.

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i Modification: F1-89-066

, JAF-SE-89-132 Rev. 2 REPLACEMENT OF SERVICE WATER PIPING j AT RBCLCW SYSTEM HEAT EXC.4ANG398 j The purpose of the modification was to replace the Service vatm1 piping downstream of the Reactor Building Closed Loop Coelity:

Water System Heat Exchangers 15E-1A, 1B, and IC.

j This modification replaced the existing carbon steel pipir.g from the outlet of the RBCLCWS heat exchangers up to and inclnling the 45 degree elbow at Elevation 317'-4" with chrome-moly piping. No rerouting of piping or revisions to the number and location of

pipe supports was required. Chrome-moly piping provides improved resistance to erosion and corrosion.

l New 150 lb. carbon steel slip-on flanges have been installed on j the heat exchanger nozzles to mitigate any possible detrimental effects on the heat exchanger due to code required post weld heat t treatment of carbon steel to chrome-moly welds.

! This modification also installed two new flanges on line 20"-WS-j 151-34 at Elevation 317'4" just downstream of the last 20" 45

! degree elbow in the Reactor Building. The installation of these flanges allows compliance with the Technical specification requirements of paragraph 3.7-C and 1.0-S for maintaining

, secondary containment integrity, if required, during the installation of this modification. Secondary containment j integrity would be maintained by the installation of a blank

between the flanges.

i The installation of the replacement piping required the isolation

! of service water to the RBCLCW heat exchangers resulting in loss of the RBCLCW system. A Temporary modification to provide

cooling water to various RBCLCW system loads that require cooling i during this outage as required. ,

[ l There are no unreviewed safety questions pursuant to 10CFR50.59. i Page 10

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Modification: F1-89-092 JAF-SE-90-009 Rev. 1 SPARE TRANSFORMER FOUNDATION The purpose of this NSE was to evaluate the potential safety I impacts associated with the installation of the foundation for the spare reserve, normal and main transformers. The scope of this NSE is limited to evaluation of the foundation and associated power feed for the transformer control box heaters and )

fans. The transformer equipment to be stored on the foundation '

are not part of this NSE. Temporary utilization of the area for storage of plant waste oils was also evaluated.

In order to prevent a pessible long forced outage, procurement of a spare reserve station transformer is planned. In the future, NYPA expects to also purchase spare normal and main transformers.

The foundation has been sized to adequately store all three (3) transformers. Since the transformers' oil reservoirs will be full during storage, the design will provide for oil containment in the unlikely event of a rupture of oil tanks. Should this occur, oil will be collected in a sump and pumped out using approved means.

The foundation is isolated from the main plant and is self-sufficient, with all electrical utilities required to maintain the reserve station transformer in storage.

It is therefore concluded that this modification does not involve an unreviewed safety question pursuant to 10CFR50.59.

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Attcchment 1 Modification: Mi-88-049 JAF-8M-90-069 Rev. 1 DISABLING OF VALVE ACTUATORS 67TCV-141 &

142 (OP) AND REMOVAL OF TEMP.

CONTROLLERS 67 TIC-141 & 142 The purpose of this modification was to disable two temperature control valve actuators for 67TCV-141 & 142, which leaves the valves in their normal fail-open position, and remove the two associated indicating temperature controllers 67 TIC-141 and 142 located in the pipe and cable tunnels. The valve actuators were obsolete and would have required total valve / actuator assembly replacement to achieve operability.

A heat balance calculation for the tunnels has shown that temperature control of the ventilation air flow is not necessary if the valves are left in an open position to permit full service water flow through the two ventilation system cooling coils 67E-11 and 14.

Calculation JAF-CALC-TBC-00839 shows that under normal plant conditions, the tunnels will maintain a minimum temperature of 40*F with an outside air temperature of -15*F.

This modification does not involve an unreviewed safety question pursuant to 10CFR50.59.

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Attcchacnt 1 Modifications M1-90-176 JAF-BE-90-084 Rev. O INSTALL BAMPLE CONNECTIONS FOR INSTRUMENT AIR SYSTEM QUALITY TESTING The purposa of this modification was to install test connections downstreau of the following root valves: 39 BAS-400, 39IAS-66, 29IAS-120, 39IAS-2179. The test locations consist of an isolation valve, and quick connect fitting. These valves shall be tagged "for air quality testing only" to prevent inadvertent i uses of instrument air. l JAF has committed to ensuring that the Instrument Air System air quality meets the Instrument Society of America standard for air quality, ISA S7.3. To meet this commitment JAF Radiological and Environmental Department personnel have issued procedure PSP-29 to test the air quality at various locations. This modification evaluates the installation of air test connections to facilitate the air testing described in PSP-29.

This modification involved the addition of tubing and valve downstream of normally closed Instrument Air System branch isolation valves. The Instrument Air System and the subject valves are Quality Assurance Category II/III non-seismic and do not perform a safety function. The location of the new tubing and components was reviewed to assure that no safety related equipment would be affected by this modification. The addition of the valve and quick connects for testing did not change the function of the air system or the subject root valves. The location of the test connections and the use as described in the test procedure PSP-29 was reviewed to ensure that the instrument air supply to other components was not affected.

There is no unreviewed safety question pursuant to 10CFR50.59.

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Attcohn nt 1 Nodification: F1-90-195 JAF-SE-90-087 Rev. 1 REPLACRNENT OF REACTOR RECIRCULATION WATER 8 AMPLE LINE CONTAINNENT ISOLATION VALVE 8 02-280V-39 AND 40 WITE 02-2AOV-39 AND 40 The purpose of this modification was to replace Reactor Recirculation Water Sample Line containment Isolation Valves 02-2SOV-39 (Line No. 1"-WH-1504-14) and 02-2SOV-40 (Line No. 3/4"-

WM-1504-14), at Penetration X-41, with new air operated valves in the existing location of the SOVs.

In addition to replacing containment isolation valves with new air operated valves that meet the LLRT criteria of Procedure ST-39B and 10CFR50, Appendix J, this modification also restored line 3/4"-WH-1504-14, (which was cut and capped in the Reactor Building) to its original design configuration so that the flow to the Crack Arrest Verification (CAV) System and the Sample Sink was restored.

All work was designated as QA Cat. I, with the exception of piping downstream of 02-2AOV-40 and air supply line to 02-2AOV-40, which were designated as QA Cat. II/III.

There is no unreviewed safety question pursuant to 10CFR50.59.

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I Attcchacnt 1 Modification: M1-91-004 JAF-8E-91-030 Rev. 1 CHEMICAL DECONTAMINATION - ADDITION ,

OF PERMANENT PIPING AND VALVE -

ADDITION OF OONCRETE BUILDING The purpose of this modification was to provide permanent piping l connections on the Reactor Water Recirculation and RWCU suction piping and a precast concrete structure to house components for chemical decontamination. The chemical decontamination process was not addressed in this modification, only the permanent modification to support the process. [

The Reactor Recirculation system including the affected branch lines is defined as QA Category I in accordance with JAF FSAR Section 12.2.3. The addition of piping and valves to RWR branch j connections will not alter or impact system operation. The newly installed valves will remain closed and flanges will remain blanked during normal plant operation.

The ultimate objective of the chemical decontamination of various systems was to reduce overall man-rem exposure. During the modification to provide chemical injection pathways to j decontaminate the piping systems, a minimum amount of work was i performed in order to maintain man rem exposures ALARA.

The addition of pipe support to each of the branch connections ,

will reduce the additional pipe stresses imposed due to addition of valve weight onto the branch piping and will ensure pipe l stresses to be less than the code allowables.  !

I To facilitate the chemical decontamination, a precast concrete l building structure has been installed adjacent to the standby gas j treatment building. This precast building was designed to. 1 provide radiation shielding and containment for the ion exchange columns and their associated equipment which are used in the decontamination process. '

Leak Testing of the new assemblies will be performed in accordance with the JAF FSAR and applicable code requirements of ANSI B31.1 (1967) and ASME Section XI (1980 Edition thru winter 1981 addenda).

The safety evaluation concluded that the modification did not involve a change in the Technical Specification or an unreviewed safety question pursuant to 10CFR50.59.

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I Attcchm0nt 1 Nodification: F1-90-171 JAF-SE-91-035 Rev. 1 REPLACEMENT OF SERVICE WATER PIPING AT RBCLCW SYSTEM HEAT EXCHANGERS j i

i This modification replaced a portion of the existing 20" carbon l steel supply header and the 12" carbon steel supply piping to the RBCLCW heat exchangers with new chrome-moly piping of the same size and schedule. The chrome-moly piping provides improved resistance to erosion and corrosion.

New 150 lb. carbon steel slip-on flanges have been installed at the heat exchanger nozzles to mitigate any possible detrimental effects on the heat exchanger due to code required post weld heat treatment of carbon steel to chrome-moly welds.

The installation of this modification does not affect any safety related components or systems. Equipment for this modification did not require environmental qualification nor vill this modification have any affect on existing environmentally qualified equipment.

There is no unreviewed safety question pursuant to 10CFR50.59.

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Att0chment 1 Modification: M1-90-233

• JAF-8E-91-050 Rev. O FEEDWATER NO2ZLE MONITORING SYSTEM The purpose of the modification is to measure the temperature of l the feedwater nozzles. Monitoring the temperature will provide I the input data to analyze the leakage flow rate of the feedwater sprayer / thermal sleeve. Excessive leakage flow past the thermal l sleeve can cause feedwater nozzle cracking due to thermal fatigue :

cycling. I The scope of the modification was to install the feedwater temperature monitoring system which consists of RTD's and a i recorder.  ;

l This modification shall remain installed for a period of five years.

Three (3) RTD's were installed on each of four (4) feedwater nozzles. They are strapped to the outside surface of the feedwater nozzles with stainless steel bands and they are located as close to the nozzle forging radius as possible. The present pipe insulation was removed to facilitate RTD installation and re-installed.

The feedwater monitoring system has been used in other BWR plants as justification for delaying the inspection requirements of NUREG 0619 (BWR Feedwater Nozzle Cracking). NRC has recognized the feedwater leakage monitoring system as a potential alternative to in-vessel inspections (NRC Docket Number 50-263).

The components have no safety function. They are not listed in the Technical specification. They are also QA Category II/III.

A material component evaluation determined that the non-nuclear grade components utilized in this modification will survive the normal temperature and radiation environments of the drywell area, for the duration of the installation stay.

JAF-CALC-FWS-00257 determined that the components and its associated devices will survive the designed seismic event and will not affect other safety related components in the drywell area.

There is no unreviewed safety question pursuant to 10CFR50.59.

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Attnehment 1 Modification: M1-91-179 JAF-SE-91-071 Rev. O ISOLATION OF ELECTRIC BAY UNIT COOLERS FOR APPENDIX R FIRE 8 This modification modified existing circuitry for system 67 unit cooler 16A & B fans in order to maintain remote operation in the event of fires in the Main Control Room or Turbine Building.

Prior to this modification, the control circuitry was tied together in a common enclosure causing concern that a fire in either location could cause short circuiting, resulting in a loss of control power to both unit coolers.

Activities accomplished included disconnecting, and sealing and sparing cables. Associated jumpers were added to maintain circuit integrity and maintain auto-start function.

This modification will resolve a 10CFR50, Appendix R/ Fire Protection issue by ensuring the 67UC-16A fans will remain operable in the event of a fire in the Turbine Building and the 67UC-16B fans will remain operable for a fire in the Main Control Room.

This design change does not constitute an unreviewed safety question pursuant to 10CFR50.59.

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i Attochm:nt 1 Modification: F1-87-061 ECW 119 IP #8 I *JAF-SE-91-113 Rev. O CONTROL ROOM ANNUNCIATOR WINDOW BRIGHTNESS This installation procedure replaced the existing type 1829 light bulbs in specific Control Room panel display assemblies with type l'

1820 light bulbs of higher wattage. The type 1820 light bulbs will provide increased luminescence of the annunciator windows for the operators. To compensate for the increased power i requirements. (13) power supplies have been installed in panel 09-43. This resulted in additional wiring in panel 09-43 and in the associated Control Room panels.

Replacement of the existing Main Control Room panel annunciator window light bulbs was necessitated as a result of the implementation of F1-87-061 IP #7 which prioritized the Main Control Room annunciators by installing colored boots on certain existing annunciator window bulbs. Along with the installation of the colored boots, new annunciator window light bulbs were installed to provide the required brightness of the color coding.

The replacement light bulbs were of increased wattage and the existing power supplies could not supply required power for the proper brightness of the booted annunciators.

The changes outlined in this installation procedure aids Control Room operators and improves the human factors aspect of the Main Control Room panels.

There is no unreviewed safety question pursuant to 10CFR50.59.

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! Attochmont 1 l Modification F1-91-269

• JAF-SE-91-122 Rev. O MAIN GENERATOR CO2 VAPORIBER REPLACEMENT j The purpose of this modification was to replace the existing

steam fired JAF main generator co2 vaporizer (76E-7) with a 65 kw electric unit. This modification also provides an independent electric energy supply to the vaporizer unit, thereby permanently separating the vaporizer from the existing auxiliary boiler system, which no longer provides steam process capability.

The scope of this modification included:

• Removal of the existing steam vaporizer unit and replacement with a new electric type unit.

+ isolation of existing steam supply piping.

+ interchanging the spare 90 amp feeder unit at 71MCC-232-OA4 with the spare size 1 starter unit at 71MCC-241-DF1, to provide power to the new vaporizer from 71MCC-241.

} + installation of new raceway and cable from 71MCC-241-DF1 to new vaporizer unit.

+

re-piping of existing CO2 piping local to the vaporizer

, unit.

There are no unreviewed safety questions pur..tuant to 10CFR50.59.

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l AttachmOnt 1 Modification N/A JAF-SE-92-006 Rev. 4 CHEMICAL TREATMENT OF SCREENWELL i FOREBAYS TO REMOVE SEBRA MUSSELS The purpose of Safety Evaluation JAF-SE-92-006 is to describe the chemical treatment method for the removal of zebra mussels from I I

the screenwell during the 1993 maintenance outage. The treatment method involves injecting the Betz Clam-Trol product in the screenwell upstream of the trash racks and allowing the screenwell and circulating water system to operate in a

" recirculation" mode. ,

1 Following the treatment for the removal of zebra mussels in the  !

screen house forebays and the circulating water system, utilizing i Betz Clamtrol product CT-1, de-activation with Betz DT-S chemical was performed to prevent discharge of active CT-1 into Lake l i

Ontario.

There are no unreviewed environmental questions and no unreviewed l safety questions pursuant to 10CFR50.59. l 1

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i Attcchm nt 1 Modification: M1-90-094

i JAF-SE-92-059 Rev. 0 ELECTRICAL OUTLETS FOR RBS WORK STATION I

, AND A WATER COOLER IN SCREENWELL AREA I

The purpose of this modification was to install electrical receptacles for the RES work bench set-up, Rad monitors and a  !

water cooler in the Screenwell area (elevation 272'0"). j i

The equipment is QA Category II/III and does not adversely affect i the safety objective and safety design bases as described in the l t

~

JAF FSAR section 8.5. The receptacles and their operation have )

no safety-related functions associated with them. The JAF  !

Technical Specifications are not affected by this modification  !

because operation of the receptacles is not important to safety.

The installed receptacles in the Screenwell area and associated components will not impact any safety-related system or affect overall plant safety, will not affect the EQ, Fire Protection, QA or Security Programs. There is no change in the FSAR and no reduction in the margin of safety as defined by the Technical Specifications.

No unreviewed safety question exists pursuant to 10CFR50.59.

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Attochacnt 1 Modification: F1-87-099 JAF-SE-92-081 Rev. 1 REACTOR WATER RECIRCULATION PLOW TRAN8MITTERS REPLACEMENT AND SPAN REDUCTION The purpose of this modification was to replace eight (8) existing Reactor Water Recirculation (RWR) Flow Transmitters 02FT-110A thru 02 FT-110H. The existing transmitters operated on a region of the square root curve which was difficult to calibrate and subject to drifting out of calibration. Other Safety Related (QA Category I) work associated with this installation included:

1) The replacement installation of eight (8) three valve manifolds associated with the transmitters.

2) Replacement of the recorder scales and chart paper for the RWR Loop Flow Recorder 02FR-163.

3) Replacement of the meter scales for RWR Loop Flow Indicators 02-159A and 02-159B.

4) Installation of additional top restraint for racks 25-23 and 25-24 to increase the factor of safety for horizontal seismic loads.

Non Safety Related (QA Category II/III) work includes:

5) Installation of two (2) Calibration Volume Chambers (CVCs e.g. water pots) 25-40 and 25-41 located in close proximity to the Flow Transmitters.

Replacing the existing RWR Flow Transmitters 02FT-110A thru 02FT-110H, complete with the associated components and work described in 1 thru 5 above reduced out-of calibration conditions and improved plant operability. Qualifications of panels and racks associated with this modification were not degraded. In addition, this modification provides for permanently installed calibration stations which reduces the need for portable equipment and minimizes set-up times, thereby reducing radiation exposures and enhancing the JAFNPP ALARA Program. These calibration stations are scheduled for installation during the 1994 Maintenance Outage.

The safety evaluation concluded that this modification does not involve an unreviewed safety question pursuant to 10CFR50.59.

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Attechm:nt 1 Modification: F1-92-109 JAF-SE-92-094 Rev. 1 EAST & WEST CABLE TUNNEL FIRE SUPPRESSION SYSTEM REPLACEMENT i l

This modification involved the removal of existing fire protection sprinkler system piping and control panels, devices, detectors and pull stations for the six (6) fire protection zones located in the East and West cable tunnels. The former dry pipe systems had been declared inoperable due to inability to provide adequate coverage.

A new wet pipe system has been installed, one (1) system per i tunnel with three (3) zones of smoke detection. This coverage j upgrades the classification to Extra Hazard Group 1 as defined by NFPA 13. In addition, changes were made to supply manifold No. 5 to assure additional flow to support the greater spray coverage for full crea suppression.

Two (2) new control panels were installed in place of the original six (6) panels drawing 125VDC power from the same source. Modifications to the control room fire protection panel were also made accordingly.

This modification provides sufficient fire suppression spray coverage for the plant to meet the design requirements of the National Fire Protection Association Standards 13 and 15 for the East & West cable tunnels.

This modification does not involve any new or unreviewed safety questions pursuant to 10CFR50.59.

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Attccha:nt 1 Modification: M1-92-292 JAF-SE-92-167 Rev. O CALIBRATION SETPOINT CHANGE FOR OFFGA8 HIGH AND LOW FLOW STRAM FLOW INSTRUMENT The purpose of this modification is to modify the instrumentation is monitoring the steam flow to the off Gas System which is designed to prevent the release of off-gas to the environment in the event of a high radioactivity level. The plant power uprate will increase the core hydrogen generation proportionally. The volume of the off-gas is reduced via recombiners, where the Hz and 02 mixture is diluted with steam to reduce the hydrogen concentration.

This modification changes the alarm and trip setpoints as follows:

Low Alarm and trip from <6069 to <6584 lb>hr.

High Alarm and trip from >7131 to >7616 lb>hr.

By changing the setpoints of the recombiner dilution steam flow instrumentation (needed for the power uprate), differential pressure switches 01-107FS-102AL and -102AH alarm and automatically isolate the off-gas system at low flow, less than 6,000 lb/hr and at high flow greater than 7,200 lb/hr.

There is no unreviewed safety question pursuant to 10CFR50.59.

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Modification: M1-89-058 ,

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• JAF-SE-92-176 Rev. O REACTOR WATER SENSING LINES BARRIER j REPLACEMENT l 1

i The purpose of this modification is to resolve a temporary modification (#92-089) that was previously installed to protect three Reactor Fater Level Sensing Lines located on the Reactor Building 326' elevation from impact damage. This modification is I intended to incorporate the temporary modification into approved plant design documents. This involved the design, documentation, evaluation and verification required for a permanent solution to this outstanding temporary modification and justify its final disposition.

The temporary modification fabricated and erected a barrier of expanded metal, 2" x 2" x 3/8" angle and plate conforming to ASTM-A-36. The structure was secured to the concrete with Hilti II KWIK bolts per IS-S-02. Steel connections used bolting materials of ASTM-A-449 and ASTM-A-194 if welding was not practical. Welding of material conformed to JAF Welding Manual and painting of completed barrier was in accordance with IS-M-01.

A calculation was performed which was in accordance with A.I.S.C.

" Manual of Steel Construction" for deadweight & seismic consideration. This calculation verified the barrier design was adequate. (Ref. JAF-CALC-NBI-00477).

This existing barrier, which is a QA Category II/III seismic class II installation, will remain in place and all applicable plant drawings will be updated to reflect the as-built configuration.

There is no safety question pursuant to 10CFR50.59.

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, Attochm:nt 1 ,

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, Modification M1-92-338 i JAF-SE-92-195 Rev. O REACTOR WATER LEVEL CONDENSING CHAMPER TEMPERATURE MONITORING SYSTEM 4 This modification installed a Reactor Water Level Condensing Chamber Temperature Monitoring System consisting of ten (10) Type i "T" thermocouples, two (2) per condensing chamber, in the Drywell Area. This system monitors differential temperature (steam vapor / condensation) to ascertain condensing chamber performance.

The condensing chambers being monitored are 02-3CCH-1, 2A, 2B, 3A and 3B, A material component evaluation has determined that the non-nuclear grade components to be utilized in this modification will survive the normal temperature, radiation and nitrogen j environments of the drywell area for the duration of the installation stay.

Data collected will be supplied to the BWR Owners Group (BWROG) research program to determine if non-condensable gases are a potential problem at JAFNPP.

I This modification does not present an unreviewed safety question pursuant to 10CFR50.59.

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l Modification M1-92-356 JAF-85-92-214 Rev. O STRUCTURAL MODIFICATION TO THE CAD BUILDING This modification installed two additional roof supporting members to seismically qualify the building structure to class I design standards from class II. I l

Installation of this modification has increased the safety I factors against the combined loads, including seismic, to that normally accepted for the drilled-in wedge type anchors used to anchor the steel structure to its foundations. The increased safety factors will enhance the strength of the building and provide additional assurance that the nitrogen storage tanks would not be affected by seismic loads on the structure.

The modification did not change the general arrangement of the structural steel in the building and does not affect any system or component.

There is no unreviewed safety question pursuant to 10CFR50.59.

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AttOchm:nt 1 Modificationt M1-90-188 JAF-BE-92-215 Rev. O ALARM TALLEY PANEL UPGRADE This modification contains ossontial safeguardo information.

There is no unroviewed safety question pursuant to 10CFR50.59.

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Attcchacnt 1 Modification M1-92-379

• JAF-SE-92-219 Rev. O REPLACEMENT OF 10MOV-39A HOTOR OPERATOR This modification replaced the existing SMB-0-25 operator on RHR Torus Cooling Isolation Valve 10MOV-39A with a new larger SMB 40 operator in order to comply with NRC Generic Letter 89-10.

10MOV-39A is a normally closed containment isolation valve and is i required to open during the Torus Cooling and Torus Spray modes l of RHR operation, t'

The replacement operator used for 10MOV-39A was originally purchased under P.O. Number S-91-16538 as a replacement operator for valve 27MOV-122 and was equipped with a spare 40 ft. lb. AC motor. The operator required a new stem nut which was compatible with the exiting valve stem and a new motor pinion and worm shaft to provide the approximate opening and closing times of the original operator. The operator spring pack was also changed by this modification.

A plate of compatible material was fabricated and welded to the l yokearm to permit mounting of the new operator.  ;

1 Wm Powell Co., the valve manufacturer, prepared Seismic Report S- l 61020 MOD which contains the seismic calculations that demonstrate '

that the modified valve meets the requirements of Specification [[::JAF-88-04|JAF-88-04]].

The replacement SMB-1-40 operator provides the additional thrust to satisfy the requirements of Generic Letter 89-10.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attachacnt 1 Modification: M1-92-407 JAF-SE-92-245 Rev. O AUXILIARY BOILER VALVE UPGRADE 87LCV-109 j The purpose of this modification is to justify the upgrade of

, check valve 87AHB-287B to an nitrogen-operated valve which opens with high electric trap level via 87LS-131. This upgrade is i necessary since the Electric Trap becomes pressurized during high l nitrogen flowrates and steam flows through 87AHB-287B and pressurizes the condensate return lines to the temporary inerting boiler. This modification is part of the AQCR 92-377 resolution

, and will obviate the need for jumper 92-313 which upgraded the l nitrogen line to 87AHB-287B from copper to stainless steel '

. tubing.

! Valve 87AHB-287B (check valve) has been replaced with an air-operated globe valve (87LCV-109). This valve will be operated with Nitrogen from the CAD system. The associated solenoid operated control valve (87SOV-109) will open 87LCV-109 in response to a high level switch activation in the Condensate Receiver Electric Trap Tank (87TK-6) and will close when this level switch deactivates. Valve 87LCV-109 will fail closed to prevent high pressure upstream of 87LCV-109 from pressurizing downstream piping and returning steam to the temporary inerting boiler.

As part of this modification, plant valves 27 CAD-57A and B and 27 CAD 58A and B will be shown as normally closed to separate QA Cat. I and II piping. These valves are safety related, seismic class I valves which supply nitrogen to operate-Auxiliary Boiler valves and will only be opened during Drywell inerting (before startup) when an operator is present in the CAD building.

There is no unreviewed safety question pursuant to 10CFR50.59.

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Attcchasnt 1 Modification: N/A JAF-SE-92-246 Rev. O WELD EXAMINATION AND POST WELD HEAT TREAT REQUIREMENTS l

The purpose of this safety evaluation is to analyze changes to i the weld examination program used to verify the pressure l integrity of piping and equipment pressure parts. These changes provide revised examination requirements for radiography, liquid penetrant testing, magnetic particle testing, ultrasonic testing i and visual examination requirements to be used following maintenance, repair, replacement and modification of existing plant systems or installation of new systems. In additicn, post weld heat treat (PWHT) requirements for P#1 carbon steel materials will be changed to meet ASME Section III criteria.

These changes will make the examination requirements consistent with those of the Inservice Inspection program currently in use at the plant. The current Inservice Inspection program conforms to ASME Section XI 1980 Edition through Winter 1981 Addenda.

The use of ASME XI-1980 Winter 81 addenda as the code of record, allows JAF to reconcile the past practice of using later editions of B31.1. This reconciliation is consistent with the requirements of ASME XI - IWA4120 and IWA7210. An engineering report, JAF-RPT-MISC-01130, details the technical justification that supports these changes.

Future weld examination requirements will be in accordance with ASME XI 1980 Edition Winter 81 Addenda and USAS B31.1-1967 Edition, A 1969 Addenda for the systems designated as ISI Class 1,2 and 3.

Future welding of Carbon Steel P#1 material may use the PWHT requirements of ASME Section III - 1992 Edition.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attcchm:nt 1 l Modifications M1-93-009 JAF-SE-93-002 Rev. 0 31MOV-RB8V-2 VALVE REPLACEMENT This modification replaced the present Crane 10"-600# motor operated pressure seal gate valve with a Powell 10"-900# motor operated pressure seal gate valve.

The existing operator was reused which has a 171.60 gear ratio and 600VAC motor. This conforms to the plant power supply and maintains an acceptable stroke time.

Replacement of 31MOV-RSSV-2 Reheater Steam Supply Valve involved cutting the 10" schedule 80 carbon steel pipe, electrical determination and retermination, installing the old operator on the replacement valve, welding the new carbon steel valve into place, radiography on all circumferential butt welds, valve functional test and leak test. Valve packing was changed to a live load configuration prior to modification testing.

The normal function and failure mode of the valve will not be changed by this modification and therefore the operation of the Moisture Separator reheater will not be affected.

This modification does not involve an unreviewed safety question pursuant to 10CFR50.59, l

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Attacharnt 1 Modification: N/A i

JAF-SE-93-004 Rev. O REACTOR WATER CLEANUP BLOWDOWN MODE I VALVE LINEUP CHANGE The purpose of this Safety Evaluation is to verify the acceptability of changing the Reactor Water Cleanup blowdown mode valve line up. The blowdown isolation valve to radwaste, 12MOV- l 57, will be opened for an extended period of time as an interim I means of clearing control room annunciator 9-4-2-5, CLN UP SYS DISCH PRESS HI OR LO, until a modification can be performed.

1 The control room annunciator is caused by a small amount of leakage past the blowdown flow control valve, 12FCV-55, that is l trapped by the two valves, 12MOV-56 and 57, creating a high I pressure condition in this volume causing the blowdown high pressure switch, 12PS-108 to alarm in the control room. This creates an undesirable " nuisance alarm" that distracts operations personnel.

This change in RWCU blowdown mode valve line up is acceptable on an interim basis provided the FCV leakage rate is monitored on a weekly basis and special condition tags are hung per WACP 10.1.2 l alerting operators of the abnormal valve line up. l It has been concluded that there is no unreviewed safety question pursuant to 10CFR50.59.

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i' AttcchaOnt 1 Modifications N/A JAF-SE-93-005 Rev. O DOWNGRADING THE QA CATEGORY 0F THE REFUELING INTERLOCK 8 FROM QA CATEGORY I TO M The purpose of this Safety Evaluation is to evaluate the acceptability of downgrading the refueling interlocks from QA )

Category I (Nuclear Safety Related) to QA Category M (Important to Safety).

The refueling interlocks include circuitry which senses the position of control rods using inputs from the Rod Position Indication System (RPIS) and restricts control rod withdrawal based on the position of the reactor mode switch and fuel handling equipment status. These interlocks provide a backup to  ;

Operating Procedures which prevent addition of positive reactivity to the core during core alterations by more than one method at a time.

Downgrading the refueling interlocks from QA Category I to M will j not:

• increase the probability of occurrence or consequences of an accident or malfunction of safety-related structures, systems or components previously evaluated in the FSAR because the refueling interlocks perform an accident preventive function which are a backup to administrative controls governing core alterations.

1 create the possibility for an accident or malfunction of safety-related structures, systems or components of a different type than previously analyzed in the FSAR.

reduce the margin of safety f i

There are no unreviewed safety q +

.lons pursuant to 10CFR50.59.

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Attechm:nt 1 Modification F1-93-017 JAF-SE-93-006 Rev. O BBGT/ TRACK BAY FIRE DOOR MODIFICATION The purpose of this modification was to identify, resolve and repair deficiencies associated with fire door 24R-272-10, the l door to the Standby Gas Treatment (SBGT) Room from the Reactor l Building Track Bay.

The deficiencies included:

- The SBGT Room did not have indication lights to inform personnel when operation of the SBGT door is permissible.

- The electric strike latching mechanism continually jammed.

This was a high maintenance item which often required replacement.

- The door did not latch in accordance with the requirements for a three (3) hour fire door as specified by NFPA-80.

The installed weather stripping did not always allow the door to soul properly.

+

The door / frame cleart nces required adjustment to meet the criteria specified by NFPA-80.

• The position indication light and alarr did not clear at the Control Room Fire Protection Panel (FPP) after the " Alarm Acknowledge button was pressed.

l In addition, new indication lights were installed in the Track Bay in close proximity to the SBGT door which will mimic the new indication lights installed in the SBGT room. Signs will be located in close proximity to the SBGT and Pass door to provide instructions for door operation.

It has been concluded that the modifications associated with door 24R-272-10 did not degrade the design basis or functions of the plant equipment and will not ravolve an unreviewed safety question pursuant to 10CFR50.59.

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Attccha:nt i Nodification: N/A JAF-SE-93-007 Rev. O TEMPORARY NODIFICATION OF HIGH PRESSURE COOLANT INJECTION SUPPRESSION POOL LEVEL INSTRUKENTATION PROCESS LINES i The purpose of this safety evaluation is to verify the acceptability of a temporary modification of the HPCI Suppression Pool Level Instrumentation Sensing Lines. This temporary modification is required to bypass a water seal created by the incorrect installation of the 23LT-201C Low Root Isolation Valve t (23HPI-949) causing an oscillating level indication during torus )

venting. Stainless steel tubing will be installed between 23LT-201C Low Side Drain Valve and 23LS-91B Vent Valve, thus bypassing the 23HPI-949 water seal. This temporary modification will provide an interim solution until maintenance can be performed to correct the valve orientation.

This temporary modification of the HPCI suppression pool level transmitter (23LT-201C) and HPCI suppression pool level switch

( 23 LS-91B) instrumentation sensing lines is acceptable on an interim basis with the following requirements:

1) the integrity of the newly installed sensing line connections is verified by a type "B" LLRT test prior to opening the 23LS-91B instrument vent;

2) the 23LT-201C indication is trended versus the 23LT-201B indication in order to predict for the possibility of accumulated condensate; i

j 3) the valve positions of the 23LT-201C low instrument

, isolation valve (shut), 23LT-201C low instrument drain valve

! (open) and the 23LS-91B instrument vent valve (open) be independently verified to assure compliance with the jumper instructions.

l' It has been concluded there is no unreviewed safety question pursuant to 10CFR50.59.

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AttcchmOnt 1 Modification: F1-91-085 JAF-SE-93-008 Rev. 1 CONTROL ROOM ACCESS MODIFICATION The purpose of this modification is to provide security tie-ins and room reconfiguration in the Control Room Work Control Center area to allow functional usage of the new door through the Control Room G-line wall.

This modification provided a security access control point at the l new door through the Control Room G-line wall. This control I point includes a card reader, palm switch and balance magnetic l switch. An existing unused zone (215) on the security access system was utilized for this control point.

The modification also included an airlock at the northeast corner of the Control Room WCC that is provided in order to maintain not positive pressure in the Control Room area.

]

The modification also provides a standard mandoor and adjacent open counter between the airlock and the existing south wall of the corridor.

This modification does not affect established critical parameters for the plant, including those contained in the FSAR, the Technical Specifications, the Security Plan, the Quality Assurance Program and the Fire Protection System.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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f Attcchment i

! Modification: M1-93-019 5

l JAF-SE-93-009 Rev. O TIP DETECTOR PROBE REPLACEMENT This modification generically approves the replacement of the i three detector probes in the traversing in-core probe (TIP) system.with one of shorter length. The shorter version was

ilesigned by the manufacturer specifically for this application at j UAF and will increase the ability of the detector probe to pass 1 through the in- coro guide tubing. This modification allows the l use of either an original detector probe or the one of shorter i length.

l This modification allowed the replacement of the existing 78mm j long detector probes and 164, foot coaxial cables with 68mm long 4

detector probes and equivalent coaxial cables of the same length.

i This decrease in the length of the detector probe will cause a 1 25% reduction in sensitivity of the TIP detector probe which is

! deemed negligible.

1 i The installation of the shorter detector probe does not increase .

I the probability of occurrence or consequences of an accident or

) malfunction of equipment important to safety. This modification j does not create the possibility of an accident or malfunction of j a different type than evaluated previously in the JAF FSAR.

j There is no reduction in the margin of safety as defined by the design basis for any Technical Specifications as a result of this modification.

I There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attcchacnt 1 Modification: M1-93-020 JAF-SE-93-010 Rev. O DRYWELL COOLING DAMPER BOV MODIFICATION This modification was performed to install and modify components where a pressure boundary failure could depressurize the safety related N2 header which is relied upon for MSIV and ADS operation following a Design Basis Accident (DBA). The ADS is required to operate 100 days post accident. l l

This modification installed a QA Category I solenoid valve in the nitrogen supply to the drywell cooling system outlet dampers.

The modification will isolate the nitrogen to the inlet dampers as well as fix the inlet dampers in their full open position.

The Core Spray and RHR Systems pneumatic operated testable check valve supply isolation valves have been changed to normally closed. The closed position of the isolation valves will provide the pressure boundary for separation between the safety related nitrogen supply system and the non-safety related operator.

l The modification to the Reactor Vessel Head Vent SOV's will not affect the operation of the Nuclear Boiler System because the design, function, failure mode and operating time of the valves has not changed by this modification.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attacha:nt 1 Modification N/A JAF-SE-93-015 Rev. O CHANGE IN REACTOR ENGINEER ORGANIEATIONAL REPORTING FROM OPERATIONS DEPARTMENT TO TECHNICAL SERVICES DEPARTMENT The purpose of this safety evaluation is to describe and evaluate a management reorganization designed to improve plant operation by moving the functional responsibility for the Reactor Engineering Group from the Operations Department to the Technical Services Department.

This change is a reassignment of position responsibilities for the Operations and Technical Services Managers. The changes do not alter the Power Authority's commitment to maintain a management structure that contributes to the safe operation and maintenance of the plant.

This plant reorganization moves the Reactor Engineering Group from the Operations Department to the Technical Services Department. No deletions or additions of responsibilities of Reactor Engineering Group personnel result from this change.

Subchapter 13.2 and Figure 13.2-7 of the James A. FitzPatrick FSAR requires revision to reflect this reorganization.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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l Attcohm0nt 1 Modification: N/A l

JAF-SE-93-016 Rev. O ACOUSTIC FISH DETERRENCE PROJECT 1993 The purpose of this project is to temporarily install and test an acoustic fish deterrence system on the offshore circulating water intake at the James A. FitzPatrick Nuclear Power Plant (JAF) from March through July 1993. The objective is to determine how effectively the system can reduce number of fish entering the intake and the number of fish subsequently impinged on the traveling water screens in the forebay.

The acoustic fish deterrence system will target the alewife, a species which comprises the majority of fish impinged on the

intake screens at JAF annually. This species is preyed upon by salmon and trout that live in Lake Ontario. The New York State Department of Environmental Conservation considers the mortality of impinged alewives to be a potentially adverse environmental impact.

This test will also be used to confirm the results of a similar test performed from March thru July 1991 under temporary MOD 92-083 and Nuclear Safety Evaluation JAF-SE-91-037. The Circulating Water system is classified as QA Category II/III. The intake structure is classified as QA Category I.

The testing of the acoustic fish deterrence system did not have an adverse environmental or safety effect and does not involve an

, unreviewed environmental question based on testing under controlled conditions, and the SPDES permits for JAF.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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i l Attcchacnt 1 Modification: N/A l JAF-SE-93-018 Rev. O REMOVAL OF FIGURE 9.10-3 FROM THE FSAR i

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! The purpose of this safety evaluation is to remove Figure 9.10-3 from the FSAR. AQCR No.92-155 noted that FSAR Figure 9.10-3 i entitled " Plant Design Water Quality Conditions" was out of date and needed revision. The response to AQCR 92-155 was to initiate appropriate evaluations for the figure and text to ensure the work performed at JAF is appropriate and correct.

FSAR section 9.10 was submitted for revision to reflect the 4

! current design and operation of the makeup water treatment system l l and the figure was deleted. In reviewing the figure and the l 1

basis for it, the figure had no relevance to the section on I i makeup water treatment. Thus, there was no value to updating it

to current practices, and the logical step was to remove it from ,

the FSAR. )

! l Figure 9.10-3 is being removed from FSAR Section 9.10 to delete I outdated and incorrect information. The correct information for original plant design is in reference 3 and 4 and should be included in DBD's as applicable. Current EPRI BWR Chemistry Guidelines are given in references 7 and 8. Plant procedures l

' incorporate these guidelines and their compliance has been l assessed by the corporate chemi-stry group. Removing Figure 9.10- i 3 will delete out of date information and has no impact on plant I operation or safety.

There is no unreviewed safety question pursuant to 10CFR50.59. i l

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Attachasnt 1 Modification: N/A J4F-SE-93-022 Rev. O TURBINE SUPERVISORY INSTRUMENTATION (TSI) SYSTEN TRIP FUNCTION DISABLING l

The purpose of this safety evaluation is to address the acceptability of disabling the main turbine and reactor feed pump turbine (RFPT) high vibration trip function to allow maintenance l and calibration activities to be performed on the TSI System l bearing vibration detectors. I Disabling the TSI high vibration trip function is acceptable for maintenance and calibration activities. Adequate instrumentation and annunciation exists for monitoring and early detection of any abnormal bearing conditions. The probability of a catastrophic component failure causing rapid increase vibrations is considered to be very low relative to the duration in which the trip function will be disabled. j 1

The components and system addressed in this safety evaluation do not perform a safety related function nor do they support other systems or components required to support a safety related function.

Disabling the TSI high vibration trip feature does not involve an unreviewed safety question pursuant to 10CFR50.59.

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! JAF-SE-93-023 Rev. O REMOVAL OF CABLE TUNNELS PRE 88URI2ATION FROM SERVICE This evaluation addresses the safety consequences of operating the plant without the pressurization of the Cable Tunnel

, Ventilation system. Furthermore, the evaluation will provide the clarification needed to allow removal of the pressurization from j FSAR section 9.9.3.4.

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Normal air flow is routed from areas of lesser to progressively greater radioactive contamination potential prior to final exhaust. By shutting down System 67, this route is not affected.

l The greatest potential radioactive contamination area exists in i the Turbine building, which is under a negative pressure. If the Cable Tunnel Ventilation system were to fail or be taken out of service for maintenance, there would be no possibility of causing, either directly or indirectly, an uncontrolled relesce of radiation in excess of 10CFR100 limits.

In summary, operation of JAF without the cable Tunnel ventilation in service to pressurize the tunnels will not adversely affect continued safe operation of the plant and does not represent a potential unreviewed safety question pursuant to 10CFR 50.59.

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r AttachmOnt 1 Modification: N/A JAF-SE-93-033 Rev. O TEMPORARY ADMINISTRATIVE CHANGE FOR THE QUALITY ASSURANCE DEPARTMENT The purpose of this safety evaluation is to describe and evaluate a temporary administrative change for the Quality Assurance Department, whereby, the administrative function of the nuclear i portion of the Quality Assurance department will report to the Executive Vice President - Nuclear Generation (EVP-NG). For issues that relate to safety and policy, the nuclear portion of the Quality Assurance Department will report to the President of the New York Power Authority. Also, for issues that involve i salary reviews, hire / fire and position assignments, the Director of Quality Assurance will report to the President. The remainder of the department (non-nuclear) will also report directly to the President. This change is being made due to a temporary vacancy in the position of Senior Vice President - Appraisal and Compliance Services.

This safety evaluation is effective until either the position of Senior Vice President Appraisal & Compliance Services is filled ,

or a new safety evaluation is prepared for permanent j organizational changes.  ;

1 The temporary administrative change does not have an affect on the overall day to day operations of the quality Assurance ,

Department nor does it alter the Power Authority's commitment to i maintain a management structure that contributes to the safe J operation and maintenance of the two Nuclear Power Plants, Indian Point #3 and James A. FitzPatrick.

There is no unreviewed safety question pursuant to 10CFR50.59.

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Attochm:nt 1 Modifications N/A JAF-SE-93-034 Rev. 0 EVALUATION OF MAEIMUM AND MINIMUM WATER LEVEL AT SCREENWELL FOR SAFE OPERATION OF CLASS I EQUIPMENT The purpose of this Safety Evaluation was to establish the basis for determining the entry conditions for Emergency Action Levels (EALs) associated with screenwell water levels. Emergency Action Levels for the declaration of Unusual Event, Alert, and Site Area Emergency for both high and low lake level conditions are determined. The entry conditions are based on assuring the ability to safely shutdown the plant and assure for the post shutdown condition the functional capabilities of plant structures, systems, and components important to safety.

The entry conditions of EALs provides early notification to l operators for any condition of lake water level changes to safe operation of QA Category I components operating in the screenwell area, l

l The EAL's have been established at 252.5' maximum and 236.5' l minimum. These maximum and minimum Screenwell Water Levels were j established based on a combination of historical recorded lake l water level on which was super positioned the effects of storm, seiche, and rainfall.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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AttOchm:nt 1 Modification: N/A F8AR Revision l

JAF-8E-93-036 Rev. 0 VERIFICATION OF INTAKE STRUCTURE DE- 1 ICING HEATERS QA CATEGORY II/III The purpose of this safety evaluation is to support a revision to l FSAR Section 12.3.7 and to verify that the intake structure de- I icing heaters do not perform a function required for the safety l function of systems, structures, or components that are needed to maintain safe shutdown, remove residual heat, control the release of radioactive material or mitigate the consequences of an accident. l FSAR Section 12.3.7 statements regarding reverse flow will also be revised to describe the extent of gate positioning required to establish reverse flow.

Based on a review of the design of the de-icing heating system and intake structure, the safety system intake flow requirements, and the documentation that makes up the FitzPatrick design bases, it is concluded that the present non-safety related, QA Category II/III classification of the de-icing heaters is appropriate.

The results of laboratory testing and analysjs indicate that frazil ice blockage will not occur during plant shutdown (i.e.

circulating water pumps out of service) with or without the intake bar rack heaters in service and that sufficient emergency service water will be available should the heaters fail during an icing event with the plant in full power operation.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attcchment 1 Modification M1-90-041 JAF-SE-93-041 Rev. O PROVIDE SHIELDING FOR SDIV LEVEL SWITCHES The purpose of this modification is to authorize the permanent installation of Temporary Modification 92-174. This Temporary

,l Modification was used to document the installation of a lead shield frame over the float cage (level pot) of the Scram ,

Discharge Instrument Volume (SDIV) level switches ( 03 LS-2 31A , B , E , l and F). This resulted in reduced radiation exposure to l technicians performing field functional tests / calibrations of I these level switches and transmitters in the SDIV protective enclosures (East and West).

This evaluation finds no conditions adverse to nuclear safety or Technical Specification design bases associated with the permanent installation of a frame to support lead shielding plates to existing support structures (level switch supports).

The lead shield structure is Q.A. Category II/III and its installation does not adversely affect systems, structure, or components associated with the safety objective and safety design bases as described in the FSAR or Technical Specifications. This modification will improve technician working conditions in a radiation area by improving access and reducing personnel

exposure.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attcchm3nt 1 Modification: N/A POT-12F JAF-SE-93-045 Rev. O PREOPERATIONAL TEST PROCEDURE POT-12F ,

IMPLEMENTATION MOD F1-90-202 l The purpose of this Nuclear Safety Evaluation is to assure that l performance of Preoperational Test Procedure POT-12F to verify operability of the Replacement RWCU pump 12P-1A will not present an unreviewed safety question.

The objectives of POT-12F are to accomplish the following:

a. Measure running and starting values of both current and voltage of the replacement pump 12P-1A motor.

b. Verify proper motor phasing and rotation of the pump.

c. Verify that the set-point of the Reactor Building cooling water temperature to pump 12P-1A is equal to 134.25'F (Ref.

6), and that the pump will trip and a high temperature alarm will be activated on panel 09-4-2.

d. Verify the operating bearing / oil temperature is less than 165'F.

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e. Perform a system flow test to develop a baseline pump curve during plant normal operation to use for monitoring pump degradation.

Pre-Operational Test Procedure POT-12F will verify operability and develop a baseline pump curve for the RWCU pump 12P-1A. The performance of POT-12F will not create any plant operating condition that was not previously evaluated as part of the design basis for the. Reactor Water Cleanup System. The components tested as part of POT-12F are non-safety related and are not required for plant shutdown or to mitigate an accident.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attccha nt 1 i l Nodification N/A l

JAF-SE-93-047 Rev. 0 FSAR SECTION 16.5 REVISION The purpose of this Nuclear Safety Evaluation was to revise Section 16.5 of the FSAR. The description of the pipe weld inspections contained in the original FSAR paragraph 16.5.2.1 was incorrect and inconsistent with the requirements contained in the original Table 1 of the JAF Piping Specification.

Nuclear Engineering and Design (NED) has determined that the requirements contained in AP-23 are correct and an FSAR revision is required. Nuclear Engineering and Design has concluded that the original FSAR text contained errors due to a poor translation of the requirements contained in Table 1 of AP-23.

FSAR paragraph 16.5.2.1 will be revised to be consistent with the requirements of the Piping Specification. Table 1 of AP-23 will be included in the FSAR and revised to correct minor typographical and omission errors.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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AttcchmOnt 1 Modification: N/A STP-20AB JAF-SE-93-049 Rev. 0 20MOV82 IN SITU DESIGN BASIS DIFFERENTIAL PRES 8URE TEST Special Test STP-20AB is performed to address requirements from NRC Generic Letter 89-10 which dictates in-situ design basis testing of motor operated valves (MOVs). This evaluation addresses testing of the Radwaste System (RDW), System 20, valve 20MOV-82 RDW Drywell Floor Drain Sump Pump Discharge Inboard Isolation Valve. This test approximates design basis differential pressure conditions across the valve to be tested, and then directs stroking of the valve. Valve differential pressure will be determined from pressure measurements taken from temporarily installed plant instrumentation. Valve operator characteristics are monitored and recorded using Liberty Technologies' " VOTES" test equipment.

Valve 20MOV-82 is the inboard containment isolation valve in the Drywell Floor Drain Sump Pump discharge Line. 20MOV-83 is open during normal reactor operations. This valve (along with 20AOV-

83) provides primary containment isolation for the Drywell Floor Drain Sump Line, closing within the specified time on a Group 2 containment isolation signal.

Perfermance of STP-20AB has no adverse affects on the Radwaste System, its components, or other plant safety related structures, systems, and components.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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a Attcchm0nt 1 Modification: N/A STP-296 3

JAF-SE-93-050 Rev. 1 23MOV-77 IN-SITU DESIGN BASIS DIFFERENTIAL PRESSURE TEST Special Test _STP-29G is being performed to address requirements I from NRC Generic Letter 89-10 which dictates in-situ design basis l testing of motor operated valves (MOVs). This evaluation i addresses testing of the Main Steam (MST), System 29, valve 1 29MOV-77 MST Inboard Line Drain Outboard Isolation Valve. This test approximates design basis differential pressure conditions across the valve to be tested, and then directs stroking of the -  !

valve. Valve differential pressure will be determined from pressure measurements taken from permanently installed plant l instrumentation. Valve operator characteristics will be monitored and recorded using Liberty Technology's " VOTES" test equipment.

Performance of STP-29G har, no adverse affects on the Main Steam System, its components, or other plant safety related structures, systems, and components. I There are no unreviewed safety questions pursuant to 10CFR50.59. I 1

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I Attcchmont 1 l

Modification: M1-92-187 JAF-SE-93-051 Rev. 0 SCREENWELL LEVEL INDICATION 1

l The purpose of this modification is to provide screenwell water l level indication to the Control Room via EPIC. The Drexelbrook l Engineering Company continuous level measurement system which I consists of one level sensor, 36LE-125 (Part #700-3-9), one level i transmitter, 36LT-125 (Part #408-8232-9) and one level indicator, l 36LI-125 (Part #601-2000-7) along with associated cable and raceway has been installed in the screenwell to provide this  ;

function. l l

Local and Control Room alarm functions are provided to alert 1 operators to a high or low water level condition due to high lake level or intake blockage. This modification is a result of DER 93-0252 corrective action due to a plant scram because of low l water level in the screenwell. l I

This modification provides continuous indication of the screenwell water level to the control room via EPIC. High and j low level alarms have been input to allow sufficient time for the operators to take the corrective actions required to maintain water level for proper operation of various cooling system pumps (NSW, CWS, ESW, RHRSW and FPS).

Failure of this equipment will not prevent operators from I monitoring screenwell water level due to other parameters and alarms that are currently in effect (screenwell water temp and differential level across the travelling screens).

This modification does not involve an unreviewed safety question i pursuant to 10CFR50.59.

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AttcchmOnt 1 l 1

Modification N/A l JAF-8E-93-053 Rev. O UBE OF CODE CASE N-356 ,

This safety evaluation analyzes the effect of adopting ASME B&PV l Code Case N-356, " Certification Period for Level III NDE Personnel Section XI, Divisions 1, 2, and 3," which extended the recertification period to five years.

This safety evaluation provides justification for updating the JAF FSAR to include the use of Code Case N-356.

Code Case N-356 has been endorsed (accepted) by the NRC for Implementation in Regulatory Guide 1.147, Rev. 9, April 1992, titled " Inservice Inspection Code Case Acceptability ASME Section XI Division 1." In addition, later editions of the Code (1983, Winter 1983 Addenda) Part IWA-2300 (a) (1) changed Level III personnel recertification period from three to five years.

Changing the recertification interval from three years to five years does not introduce any adverse effects to the examination process or Level III personnel qualification requirements since these areas are not changed. Extending the recortification period to five years also does not affect the Level III personnel's ability to perform their function, reduce their performance standard or degrade the quality of their work. This change is administrative in nature rather than technical and reflects practices accepted by the NRC and Nuclear Industry.

This change does not involve an unreviewed safety question pursuant to 10CFR50.59.

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Attechasnt 1 Modification: N/A Jumper 93-40 JAF-SE-93-054 Rev. O TEMPOR)RY COOLING WATER SUPPLY VIA TEMPORARY PUNP INSTALLED IN THE INTAKE TO CONDENSATE PUNP MOTOR COOLER AND AIR COMPRESSOR DURING SERVICE WATER OUTAGE The purpose of this temporary modification is to provide a temporary supply of cooling water to the air compressors, 39AC-1A,B,C for plant outage periods when turbine building closed loop cooling system (TBC) is out of service. The cooling water is to be supplied from a temporary pump installed in the intake between the trash rack and the travelling screen. The plant will be in the cold shutdown condition for the duration of this temporary modification.

In addition to the above, this temporary modification will provide temporary cooling water to a condensate pump motor (33P-8A or B or C) via the same flow path as described above.

A backup pump will be installed in the intake next to the primary pump. This pump will duplicate the hose configuration of the primary pump all the way to the air compressors and the condensate pump in order to prevent a singlo hose rupture or a failure of the primary pump from causing a lous of cooling.

The use of a temporary supply to cool the above TBC loads results in system operation of the TBCLC, water treatment, breathing, instrument and service air system, and condensate system different than described in the FSAR.

This temporary mod, will not affect any safety related systems, or affect overall plant safety. The plant will be in the cold shutdown condition for the duration of the temporary modification.

There is no unroviewed safety question pursuant to 10CFR50.59.

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AttochmOnt 1 i Modification N/A Jumper 93-11 )

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JAF-85-93-060 Rev. 1 REMOVAL OF VARIOUS MAIN STEAM RADIATION 1 4

MONITOR OUTPUTS TO RECORDER 17RR-252 AND l EPIC COMPUTER FOR TROUBLESHOOTING The purpose of this evaluation is to provide justification for

! disconnecting various cables providing the recorder (17RR-252) l and EPlc Computer signals from the Main Steam Line Radiation 4

Monitors (MSLRM) (17RM-251A,B,C, or D). This will allow technicians to isolate the cause of intermittent downscale ala rms .

This Safety Evaluation is being written for the following reasons: 1) I&C Departments ICSO-12 Generic Troubleshooting and Maintenance Procedure cannot be used for disconnecting the cables because it would require inserting a trip (half scram) into the affected RPS channel (s). (It is desired to maintain the monitors operable allowing work to be performed without inserting a half scram), and 2) WACP-10.1.3, Control of Jumpers, Lifted Leads, and Temporary Modifications must be used to document the

~

disconnection and Part IV-4 requires a Nuclear Safety Evaluation because it will result in operation different than described in the FSAR.

The disconnection of various inputs to the MSLRM recorder 17RR-252 and/or EPIC computer does not involve any conflicts with the FSAR, Technical Specifications, or licensing commitments. It will not affect the operation of the MSLRM's local (drawer) indication or trip functions. Since the recorder and EPIC provide indication functions only, in addition to that available at the front panel monitor, there are no effects on the safety functions of the system.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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l Att0chmOnt 1 Modification: N/A TOP-153 JAF-SE-93-066 Rev. 4 TEMPORARY COOLING WATER SUPPLY TO THE CRD PUMP COOLERS, DRYWELL COOLERS, AND  !

FUEL POOL HEAT EXCHANGERS l

! The purpose of this temporary installation is to provide temporary cooling to the CRD pump coolers ( 03 P-16 A/B) , drywell coolers ( 68 E-3 A/B/ C/D) , and spent fuel pool coolers (19E-1A/B) .

i The reactor building cooling water heat exchangers (15E-1A/ B/C) will be unavailable during the 1993 Fall maintenance outage due to a service water system modification. The required cooling water will be supplied by a temporary leased pump / chiller unit located outside the reactor building.

i The RBCLC system lineup to support operation of this temporary modification is described in TOP-153 (reference 10).

This temporary installation does not adversely affect any safety related systems, or affect overall plant safety.

l IE Bulletin 80-10 (Contamination of Nonradioactive System and i Resulting Potential for Unmonitored, Uncontrolled Release to Environment) has been reviewed for impact of this temporary l' modification. All actions specified in the Bulletin relative to identification of potentially contaminated systems, monitoring and sampling programs, and actions to take in the event of system contamination have been appropriately addressed.

l The temporary installation has no adverse effects on the ESW system to provide cooling water to safety related loads (i.e., 1 3

EDG's) during the outage. Defeating the ESW lock-out matrix is l J routinely performed during RBCLC system shutdown and is

] adequately addressed in existing plant procedures.

There is no unreviewed safety question pursuant to 10CFR50.59.

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Attcchm:nt 1 Modification: M1-92-071 JAF-SE-93-067 Rev. O HPCI TURBINE AREA LIGHTING UPGRADE The purpose of this mellfication is to install adequate, seismically mounted, lighting in the HPCI turbine room.

On the southern wall of the HPCI turbine room (2) two 200W pendant mounted incandescent lamp lighting fixtures were replaced by (2) two 175W metal halide fixtures.

On the northern (Torus) wall of the HPCI turbine room (2) two 200W pendant mounted incandescent lamp lighting fixtures were replaced by (3) three 175W metal Halide fixtures.

This modification by utilizing existing electrical circuitry and the new type of lighting fixtures has improved lighting conditions in the HPCI turbine area. The new lighting fixtures have reduced required wattage and provide increased illumination with a minimum change in installation.

This modification does not involve an unreviewed safety question pursuant to 10CFR50.59.

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l l l Attachment 1 I

Modification

F1-93-075 JAF-SE-93-072 Rev. 2 REACTOR VESSEL WATER LEVEL BACKFILL

) MODIFICATION i

i The purpose of this modification is to install a backfill system

for the reactor vessel water level instrumentation reference I legs. This modification is in response to USNRC Generic Letter i 92-04 and Bulletin 93-03, which describe the potential for inaccuracies in the reactor vessel level monitoring / indication  ;

l system at BWR plants such as JAFNPP. Bulletin 93-03 notified

holders of BWR operating licenses that they should impleraent modifications to the Reactor Water Level Instrumentation system l during the first cold shutdown after July 30, 1993, to ensure j that the water level system design provides high reliability.

i l This modification installs a backfill system for all five (5) i reactor vessel water level reference legs. The backfill fluid is provided from the CRD charging header. One CRD tie-in point branches out to supply the five (5) reference legs for condensing chambers 1, 2A, 2B, 3A and 3B. Each of the individual backfill legs are comprised of the following components:

• Turbine Flowmeter Manual Isolation Valves

+ Flow Metering Valves

. Check Valves

+ Filters

• Vents and Test Taps

• Flow Indicating Transmitter (providing local indication)

+ Hoffman Cabinets Electrical Components (Cable, terminal blocks, etc.)

The implementation of this modification does not involve an unreviewed safety question pursuant to 10CFR50.59.

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Nodification

N/A l

l JAF-85-93-073 Rev. O PLACEMENT AND USE OF THE HAIBIN STORAGE l CONTAINER FOR RADIOLOGICALLY CONTANINATED EAEARDOUS CHEMICAL j MATERIALS

) This nuclear safety evaluation is the basis for determining

! whether the addition and subsequent use of a HAZBIN chemical

storage and containment structure at the James A. FitzPatrick i Nuclear Power Plant (JAFNPP) ices or does not involve an unreviewed safety question. I The HAZBIN containment structure will be used to store mixed waste on site. The mixed waste is classified as radioactively I contaminated hazardoas chemicals and materials.

! I I I The radioactive material stored within steel containers is located within the protected area. Radiation exposure to the 3

public and workers is being controlled by securing the material, j surveillance, labeling and by situating the material'within the l 1

protected area of the site. The necessary procedural controls are being developed to ensure the use of the HAZBIN structure would not cause excessive amounts of radioactive material or flammable material to be stored in the area. An uncontrolled release to tho environment or inadvertent release of radioactive

! material is unlikely due to the structures inherent design. The radiation level at the boundary of the HAZBIN container is limited to 0.5 mR/hr or less.

l The structure will be isolated from incompatible wastes, it

maintains the required volume for the containment of any spills, l is secured from unauthorized use, and will be positioned in an

area that has minimum site traffic until it's ultimate disposal.

t i The storage of low level radioactively contaminated hazardous

! materials within a HAZBIN on-site does not constitute an l unreviewed safety question pursuant to 10CFR50.59.

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Modification

N/A POT-02W JAF-SE-93-074 Rev. O PREOPERATIONAL TEST FOR REACTOR VE8SEL i

LEVEL BACKFILL MODIFICATION (F1-93-075) I l

The purpose of this Nuclear Safety Evaluation (NSE) is to '

demonstrate that preoperational test POT-02M will not affect plant safety.

This preoperational test has been divided into four (4) portions.

The first portion of the test placed all five backfill panels into service and induced transients into the CRD and backfill systems while the plant was in a cold condition to verify that no adverse operational transients would be generated different than shown in the model.

The second portion of this test monitored and recorded plant level, presscra and backfill flow rates during plant startup.

Thir. data will be utilized to determine the effect of the bacl: fill system on plant level and pressure during Control Rod movement and other transients caused by startup.

The third portion of this test induced limited transients into the CRD and backfill systems while the plant was at approximately 100% power to determine the system response to these transients.

Again plant level, pressure and backfill flow rates were monitored and recorded.

The final portion of the test which is now in progress, will monitor and record plant level, pressure and backfill flow rates for approximately six (6) months following 100% power testing.

This data will provide added assurance of system reliability.

This preoperational test is being performed to ensure that the Reactor Vessel Level backfill system installed as Modification F1-93-075 is operating properly and will provide validation of the model, confirming the results of other transients either too difficult or impossible to test (i.e., DBA large break LOCA, MSIV closure with SCRAM and SCRAM at full power).

The performance of this test does not constitute an unreviewed safety question pursuant to 10CFR50.59.

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1 2 Attachment 1 i

! Modification: N/A

)

JA?-SE-93-076 Rev. 0 OPERATING THE OFF GA8 DRYING TOWER, 01-i 107D-6A WITH REDUCED TOWER REATER j CAPACITY i

1 The purpose of this temporary modification is to allow the continued operation of the off-gas air dryers with defective "A" tower heaters jumpered out due to a ground.

4 j This temporary modification disables an indeterminate amount-of

} the "A" off-gas drying tower heating capacity. It increases the i drying tower heating cycle time as necessary to compensate for j the reduced heating capacity. This will allow the dryers to i perform their design function of removing moisture from off-gas in turn allowing the charcoal beds to remain on service ,

maintaining the plant gaseous effluents ALARA.

i There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attcchasnt 1 Modification: N/A JAF-SE-93-077 Rev. O CHAEGE IN COMPUTER MANAGER ORGANISATIONAL REPORTING FROM GENERAL MP/4AGER, SUPPORT SERVICES TO VICE l PA38IDENT-MIS The purpose of this safety evaluation is to describe and evaluate a management reorganization which affects subchapter 13.2 of the James A. FitzPatrick FSAR. This reorganization is designed to improve and standardize plant computer systems operation by changing the functional responsibility for the computer Manager of the Management Information Systems (MIS) Department from the General Manager - Support Services (GMSS) to the Vice President -

MIS.

This change is a reassignment of position responsibilities for the General Manager Support Services and the corporate MIS Department. The change does not alter the Authority's commitment to maintain a management structure that contributes to the safe operation of the plant.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attccha nt 1 )

l Modification N/A POT-99N JAF-8E-93-078 Rev. 0 SUPPORT AND ADMINISTRATION FACILITY SECURITY MODIFICATION (DEMOLITION)

(MODIFICATION F1-90-013)

Uhe purpose of the pra-operational testing of the modification to the security and lighting systems is to verify the operability of the reconfigured perimeter such that the Support and Administration Fac).11ty Modification No. F1-90-013 is permanently located within the protected area. Verification of operability includes testing of the perimeter intrusion detection zones which were modified, reconnection of existing IR zones in the Security Building and relocation of a light fixture.

POT-99N pre-operational test verifies the operability of the reconfigured security system in order for the Support and Administration Facility to be permanently placed within the protected area.

This pre-operational test does not result in any unreviewed safety questions pursuant to 10CFR50.59.

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! I j Attachment 1 i

Modification: N/A T8T-27 l

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JAF-SE-93-085 Rev. O FUEL POOL WATER TEMPERATURE RISE TEST l The purpose of the safety evaluation is to address the j acceptability of securing Reactor Building Closed Loop Cooling l j (RBCLC) flow to the Spent Fuel Pool Coolers in order to conduct Temporary Surveillance Test Procedure TST-27 (Fuel Pool Water Temperature Rise Test. This test is being conducted to determine )

{ the rate of fuel pool temperature rise with no cooling.

{ During the Fall 1993 Maintenance Outage, a temporary chiller unit

} will be used to provide chilled water cooling to the CRD pumps, l drywell coolers, and fuel pool coolers using the RBCLC system j piping. In the unlikely event of a loss of fuel pool cooling, it

is desirable to know the rate of fuel pool temperature rise j (

• F/hr) and the corresponding duration in order to plan j appropriate actions to restore cooling before exceeding fuel pool temperature limits.

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Performing this special test is acceptable for its intended

purpose. Adequate precautions are identified in the test

! procedure along with existing alarms to alert the Operators of 1 excessive fuel pool temperatures. This special test will not j reduce the margin of safety as defined in the basis for Technical

Specification.

This test does not constitute an unreviewed safety question

! pursuant to 10CFR50.59.

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Attcchasnt 1 '

Modification: M1-92-117 JAF-SE-93-089 Rev. 1 CORE SPRAY MINIMUM FLOW LINE PIPING AMD VALVE REPLACEMENT This modification replaced approximately five feet of 3" schedule 40 carbon steel ASTM A106 GR. B pipe with ASTM A376 TP316 "ainless steel pipe which is more resistant to cavitation. No tarouting of piping or revision to the number and location of i pipe supports was required. 14RO-27A will be relocated farther upstream of 14MOV-5A and 5B to decrease future cavitation damage to the motor operated valve internals.

Also, a Pacific 3" 300# carbon steel gate _ valve 14MOV-SA, weighing approximately 150# will be replaced with a 3" Borg Warner 300# stainless steel gate valves weighing 180#. The existing motor operator will be reused. This replacement is necessary because the existing Pacific gate valve was damaged beyond repair. The replacement valve is adequate for use as a core spray minimum flow isolation valve. ,

i Valve 14MOV5B was repaired as an option to replacement.

This piping replacement has been designed in accordance with ANSI B31.1 (2967 edition) and was installed and inspected in accordance with approved plant procedures.

The JAF Technical Specification has been revised to exempt core spray minimum flow branch piping from 10CFR50, Appendix J, Type A, B, or C Test, during the 1993 Maintenance Outage. JAF Technical Specification Amendment #196 has been issued. PCILRT of this piping will be performed prior to restart from the 1994 Refuel Outage.

This modification required pre-implementation review by the NRC because of a change to plant technical specifications but does not involve an unreviewed safety question pursuant to 10CFR50.59.

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I Attcchacnt 1 Modification: N/A STP-14F/14G j JAF-85-93-091 Rev. 0 14MOV5A, -5B IN SITU DESIGN BASIS DIFFERENTIAL PRES 8URE TEST Special Test STP-14F and STP-14G are being performed to address requirements from NRC Generic Letter 89-10 which dictate in-situ design basis testing of motor operated valves (MOVs). This evaluation addresses testing of the Core Spray (CS), System 14, valves 14MOV-5A " CSP Pump A Minimum Flow Isolation Valve" and 14MOV-5B " CSP Pump B Minimum Flow Isolation Valve". This test i

! approximates design basis differential pressure conditions across the valves to be tested, and then directs stroking of the valves.

Valve differential pressure will be determined from pressure measurements taken from permanent and temporarily installed plant instrumentation.

This Nuclear Safety Evaluation allows performance of differential pressure testing under flow conditions for valves 14MOV-5A and 14MOV-5B. Valve operator characteristics will be monitored and recorded using Liberty Technologies' " VOTES" test equipment.

I Performance of STP-14F and STP-14G has no adverse affects on the Core Spray System, its components, or other plant safety related structures, systems, and components. STP-14F and STP-14G may be preformed in any reactor operational mode when all required Technical Specifications are met.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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Attccha:nt i Hodifications i

j JAF-SE-93-092 Rev. 4 LIMITED OPERhTION OF REACTOR BUILDING CRANE The purpose of this safety evaluation is to address the safety

! concerns associated with continued use of the Reactor Building

< crane while having a mechanical deficiency in the bridge drive.

The scope of this evaluation is for the limited use of the crane for high priority tasks such as the spent fuel pool cleanup operation. Repair of the crane will be performed after j replacement parts are available.

The east-west travel of the Reactor Building Crane is controlled alone by the north drive wheel due to a mechanical drive deficiency in the south drive wheel assembly. The safety design basis of the crane pertains to its load carrying capacity, which is not reduced, nor jeopardized by the deficiency in the drive mechanism. The crane bridge will be restricted to its lowest speed (s) such that the effects of bridge skew or potential brake loss are negligible. Approved procedures address all phases of the crane use, and only qualified NYPA personnel operate the crane.

In conclusion, the limited use of the crane does not constitute an unreviewed safety question as defined by 10CFR 50.59 or involve significant hazards as defined in 10CFR50.92.

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Modification: N/A l 1

! JAF-SE-93-093 Rev. 0 INCREASED STROKE TIME LIMIT EVALUATION l FOR 13MOV-27  !

! The purpose of this safety evaluation is to analyze the

< acceptability of increasing the 13MOV-27 closed stroke 5 second

} limit and to set a new operational limit. This evaluation will j be used to revise all design documents which state a closing time 2

for this valve. The new operational limit has been placed at I j 6.25 seconds.

l 1 i This increase in 13MOV-27 stroke time and limit is acceptable because the effects on system operations has been evaluated.

Also, the containment isolation function of this valve is via remote manual operation thus valve closure is dependent on  !

operator reaction time.

j It has been concluded that increasing the stroke time of 13MOV-27 i to 6.25 seconds does not constitute an unreviewed safety question

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AttachmOnt 1 Modification: N/A l

JAF-8E-93-094 Rev. 0 USE OF THE VACUUM DRYING SYSTEM FOR l REMOVING WATER FROM THE TN-RAM SHIPPING '

CASK  ;

This safety evaluation documents that the use of the Vacuum '

Drying System (VDS) on the Refueling Floor (Reactor Building l Elevation 369) as described by JAF-1006, " Procedure For Handling )

and Loading the TN-RAM Shipping Cask" does not constitute an '

unreviewed safety question. l Irradiated material is to be shipped off-site to the disposal facility in several TN-RAM shipping casks. In order to meet the  ;

federal regulations for use and transportation of the cask and '

the low-level waste disposal requirements, the VDS will be used.

The VDS is a portable, self-contained, free-standing auxiliary vacuum system used to remove residual moisture from a cask cavity by vacuum drying. In order to meet the liquid requirements, the cask has to be drained and dried. Residual water is evaporated by the VDS. A system diagram is provided in the procedure JAF-1006. The system is not interconnected with any other plant systems, therefore no system design changes will occur.

The operation of the Vacuum Drying System does not conflict with the description of the Radioactive Solid Waste system as stated in the FSAR, nor result in changes to the Technical Specifications, the Process Control Program, nor impact on any safety related or environmentally qualified structures, systems or components.

There is no unreviewed safety question pursuant to 10CFR50.59.

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Attachm0nt'1 Modification: M1-93-140 l l

JAF-8E-93-097 Rev. O RP8 ISOLATION RESISTORS RELOCATION AND WIRES REROUTING IN 09-15 AND 09-17 l 4 PANELS This modification changed the physical arrangement of the RPS isolation resistors (05A-R3A through H) in panels 09-15 and 09-1

17. This corrected a deficiency in the GE circuit design and l

meets the current licensing basis of IEEE-279 and FSAR Sec. 7.2.2 such that no single failure within the RPS prevents proper RPS action. In addition, the power wiring between the scram 1 contactors for rod groups 3 and 4 was also modified.

This change to the physical arrangement for the second set of resistors does not increase the probability of occurrence of consequences of an accident or malfunction of equipment important to safety. In addition, it does not create the possibility of 3

an accident or malfunction of a different type, nor does it reduce the margin of safety as defined in the basis for any Technical Specification.

There are no unreviewed safety questions pursuant to 10CFR50.59.

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i Attcchacnt 1 Modification: M1-93-141 JAF-SE-93-100 Rev. O TEMPORARY DISABLING OF THE AUTOMATIC START CIRCUIT OF CONTROL ROOM EMBRGENCY SUPPLY FAN 8 70FN-6A AND 70FN-6B.

This modification involved the temporarily disabling of the auto-start circuit of the emergency supply fans 70FN-6A and 70FN-6B and running both emergency fans simultaneously during a LOCA.

The automatic start feature for these fans will be restored upon resolution of a separation concern.

As both fans will be running simultaneously during a LOCA, the i auto-start of the standby fan is not required. Running of two emergency fans will result in higher flow through the emergency ventilation system components. The components were determined to be adequate to handle the increased flow.

Radiological doses to the Control Room operators due to increased flow were evaluated and were determined to be below 10CFR50, Appendix A, Criterion 19 limits. Therefore, temporarily disabling the auto-start circuit for emergency fans is

acceptable.

There is no unreviewed safety question pursuant to 10CFR50.59.

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Modification

N/A JAF-8E-93-104 Rev. O EVALUATION OF 27MOV-122 AND 27MOV-123 l CLOBE STROKE TIMES i This r.afety evaluation is written to document changes to the list of containment isolation valves in the Final Safety Analysis Report (FSAR) section 7.3 and Administrative Procedure (AP) -

01.04. Specifically, the Drywell Exhaust Bypass Inboard Isolation valve (27MOV-122) and Torus Exhaust Bypass Outboard Isolation valve (27MOV-123) will be listed as having a maximum closing time of 8 seconds instead of 5 seconds to account for actual closing times under degraded voltage conditions.

The 27MOV-122 and 27MOV-123 maximum close stroke times of 8 i

seconds are acceptable because: 1) the radiological consequences from the longer closure times are bounded by the

! accident analysis performed for a postulated LOCA during venting through the larger 24" (27AOV-113/ 27AOV-114) and 20" (27AOV-117/

27AOV-118) containment isolation valves; and 2) the l pressurization effects on the SGT system from the longer closure time are bounded by the analyses performed for a postulated LOCA during venting through the larger containment isolation valves.

This' modification does not create an unreviewed safety question

! pursuant to 10CFR50.59.

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f AttccanOnt 1 Modification N/A STP-466 JAF-SE-93-105 Rev. 0 SPECIAL TEST TO DETERMIME E8W SYSTEM RESPONSB RESULTING FROM THROTTLING 46E8W-5A/B/C/D The purpose of this special test is to collect ESW system hydraulic data to evaluate system response to throttling the Emergency Diesel Generator Jacket Water Cooler outlet isolation valves, 4 6ESW-5A/ B/C/ D. These valves are shown as " locked open" on FSAR Figure 9.7-2.

Spurious ESW strainer differential pressure alarms are commonly received whenever the system is operated with the ESW bypass isolation valves (4 6MOV-102 A/B) open. Data gathered during this test will aid in the formation of options to eliminate the spurious alarms.

The STP was performed following the completion of JAF modification S1-92-136. This modification replaced the existing cooler isolation valves with new Neles-Jamesbury butterfly valves. These Neles-Jamesbury valves possess more desirable throttling characteristics than their predecessors.

The performance of the STP is acceptable because it will not prevent the ESW system from performing its safety function. This is true because all ESW safety related loads will be receiving more flow than required by the FSAR during the performance of the special test.

Therefore, it has been concluded that performing the STP does not constitute an unreviewed safety question pursuant to 10CFR50.59.

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AttCchmCnt 1 Modification: N/A JAF-SE-93-110 Rev. 1 TECHNICAL BPECIFICATION 3.8 MISCELLANEOUS RADIOACTIVE MATERIAL 8 SOURCES PROPOSED CHANGE The purpose of this NSE is to ensure that a change from the current JAF Technical Specification (TS) 3.8, " Miscellaneous )

Radioactive Materials Sources" to adopt the Limiting Condition i for operation for the " Sealed Source Contamination" TS in NUREG 0123, Standard Technical Specification (STS) for General Electric Boiling Water Reactors; BRR/5 (Revision 3 1980) does not '

constitute an unreviewed safety question.

The current TS for leak testing radioactive sources inappropriately used 10CFR 30.71 Schedule B as the basis for exempting sources for the leak testing requirements. 10CFR 30.71 Schedule B lists those individual source activities that are exempt from requiring a radioactive materials byproduct license.

By citing part 30.71, the JAF TS was more restrictive than the STS.

A change from the current JAF Technical Specification (TS) 3.8,

" Miscellaneous Radioactive Materials Sources" to adopt the Limiting Condition for Operation from the " Sealed Source Contamination" Technical Specification in NUREG 0123, Standard Technical Specification (STS) for General Electric Boiling Water Reactors; BWR/5 (Revision 3 1980) does not constitute an unreviewed safety question pursuant to 10CFR50.59, 1

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