JAFP-83-0513, Annual Summary of Plant Mods,Changes & Experiments for 1982

Annual Summary of Plant Mods,Changes & Experiments for 1982

ML20071J947
Person / Time
Site:	FitzPatrick
Issue date:	05/13/1983
From:	Corbin McNeil POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Allan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
JAFP-83-0513, JAFP-83-513, NUDOCS 8305270189
Download: ML20071J947 (23)

Text

{{#Wiki_filter:F' J mes A. FitzP trick l, Nucle r Pow r Pl:nt RO. Box 41 Lycoming, New York 13093 315 342.3840

1. > NewYorkPower

& Authority %?p%.$ United States Nuclear Regulatory Commission Region I, 631 Park Avenue King of Prussia, Pennsylvania Attention: James M. Allan Acting Director

SUBJECT:

DOCKET 50-331-ANNUAL

SUMMARY

OF JAFNPP PLANT MODIFICATIONS, CHANGES AND EXPERIMENTS FOR 1982

Enclosure:

(a) Summary of Modifications, Changes and Tests Implemented at JAFNPP During 1982

Dear Sir:

Enclosed for your review is a Summary of Modifications, Changes and Tests implemented at JAFNPP during 1982 for your review in accordance with 10CFR50.59 requirements. Modifications, changes, experiments and tests that were completed in 1982 during the 1981-82 refueling outage were submitted in the report sent last year. If you have any questions concerning this report, please contact Mr. Victor M. Walz at (315) 342-3840, extension 265. Very truly yours, CORBIN A. MCNEIL JR. RESIDENT MANAGER CAM:KK:nvw Enclosure CC: C. Brown, JAF R. Burns, WPO J. Gray, WPO K. Kilpeck, JA F Document Control Center R. Baker, JAF R. Converse, JAF v 8305270189 830513 PDR ADOCK 05000333 R PDR L

~ _ 4

SUMMARY

OF MODIFICATIONS,-CHANGES AND TESTS IMPLEMENTED AT JAFNPP DURING 1982 j _ JA F-SE-81-005, Plant Modific ation F1-80-017 This modification consisted of installing a filtered. ventilation l system to provide filtered fresh air to the Technical Support Center via a new filter train and the existing HVAC System. The filter train, consisting of a heate r, prefilter, HEPA, and charcoal filters and a 3000 SCFM fan will be manually activated by a selector switch on local panel HV-6 at elevation 300'-0" in the Administration Building. Sequentially, the system's outside 1 air intake ( 0. A. I. ) duct ' damper HDD-106 and return duct damper A00-173 will open, while the 0.A.I. duct damper A0D-171 on system AHU-4 and the exhaust duct damper A00-146 will close. The intent of this controlled action is to allow fan FN-5 of system AHU-4 to totally circulate the air within its pressure boundary, while the fan on the filter system will add approximately 3000 CFM of filtered air to the return duct so as to give a positive pressure differential of 1/8" water gauge in the boundaries within thc l 1st, 2nd, and 3rd floors of the Administration Building, relative to the surrounding areas. i Upon actuation of the selector switch, 3 new additional dampers shall also close isolating the shops, office, stores, and first aid room, which are outside the TSC boundaries. To maintain the prescribed areas et this design pressure, it is 3 essential that FN-20 and FN-21 (r whaust and return system) in the l administration building be menually shut off when the filter system has been put into operation. Finally, to ensure that a positive pressure is maintained within the boundaries of the TSC, a complete verification of the j adequacy of all seals around doors, duct, and miscellaneous l penetrations on the pressure boundary was performed. l l JAF-SE-81-018, Plant Modification F1-81-011 l This modification consisted of installing two (2) fusible-linked counterbalanced door closing mechanisms on the Reactor Core ( Isolation Cooling (RCIC) enclosure doors. The modification is to relieve pressurc transients from inside the RCIC enclosure preventing damage to the structure,-if a.high energy line break should occur insid e. t i Page J of 22

The doors, which swing outward from the enclosure, will be held in the open position by a counterbalanced closing device with fusible links on the outside and inside of the enclosure. With this arrangement, when the temperature reaches 165*F. due to a fire either inside or outside the enclosure the heat generated will cause the fusible links to open, and the fire door will close to maintain the three hour fire integrity of the enclosure. Should a line break occur, the 165*F temperature would also be reached, however, the pressure inside the enclosure will hold th e doors open until the peak pressure transient is relieved. JAF-SE-81-029, Plant Modification F1-80-014 This modification consisted of installing three noble gas monitoring units connected in line with existing ef' fluent monitors. One unit each is connected in line with the turbine building exhaust s ampler, the radwaste building exhaust sampler and the main stack ef fluent monitor. Each monitoring unit contains two redundant detectors. Associated with each detector is a meter type readout module in t he main control room panel 09-2, having a range of 10-1 to 107 Mr/Hr. In addition to meter readouts, these modules supply digital outputs fo r annunciation of failure, high radiation, and h igh-high radiation. They also supply analog outputs for trend recording and computer logging. Th e trend recording is accomplished by three two-pen recorders mounted adjacent to the readout modules on panel 09-2. Power for the above described equipment is to be supplied from the AC Uninterruptible Power Supply (ACUPS) panel 2. This power is provided to maintain continuous operation during power transients. This modification was implemented to conform to NUREG-0737, Section II.F.1 attachment 1 which states: " Noble gas ef fluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions. Multiple monitors are considered necessary to cover the ranges of interest. (1) Noble gas effluent monitors with an upper range 5 micro Ci/cc (Xe-133) are considered to capacity of 10 be practical and should be installed in all operating plants. l l Page 2 of 22 l ]

(2) Noble gas effluent monitoring shall be provided for th e total range of concentration extending from normal condition (as low as reasonably achievable-(ALARA) concentrations) to.a maximum of 105 micro Ci/cc ( X e-13 3) ). Multiple monitors are considered to be necessary to cover the ranges of interest. The addition of the.high range effluent monitors will not af fec t the operation of the existing monitors nor will it have any impact on. the existing -FSAR statement s. The high range monitors simply operate.in conjunction with the existing monitors to extend the range of measurement capability..The descriptive sections of paragraphs 7.12.3 (Main Stack Radiation Monitoring Subsystem) and 7.12.5 (Ventilation Radiation Monitoring Sub-system) -are to be updated to reflect the addition of the high range monitors. Installation of the new effluent monitors will impact limiting conditions in the Technical Specifications - for operation, as the existing monitors must be shutdown in order to tie in the new monitors. The installation sequence.has been reviewed for impact.on the operation of the plant. No unreviewed safety question has been presented, and the probability of an accident or malfunction has .not been created. JAF-SE-81-035 Work consisted of lifting and continuous positioning of a lead lined personnel cage used while cleaning and hydrolasing the reactor internals pit. The total maximum load lifted was less 1 than 1 ton and the rigging capacity was certified to exceed 1 ton. l This work was performed cautiously, with preventive measures and administrative controls to preclude damage to any existing-equipment and to ensure relative safety for personnel performing the work. [ ( Due to the minor scope of the maintenance activity and the l previously mentioned precautions and administrative controls, I this maintenance task does not result in an unreviewed safety l question. i l Page 3 of 22 I

d JAF-SE-81-037 Work consisted of several lifts to implement the spent fuel re-racking operation. The Power Authority acquired the services of the TERA Corporation to evaluate the heavy lifts associated with this operation. The'ir evaluation indicated that measures equivalent to those required by NUREG 0612 were taken and that accordingly, adequate load handling reliability was assured. This work was performed cautiously, with preventive measures and administrative controls to preclude damage to any existing equipment and to ensure relative safety for personnel performing the work. Due to the minor scope of the maintenance activity and the previusly mentioned precautions and administrative controls this maintenance task does not result in an unreviewed safety question. JAF-SE-81-043 Work consisted of bringing a new gantry crane into the Reactor Building via -the Track Bay to th e Hoist Well, then lifting it up to the Refuel Floor, moving it to the Dryer-Separator Pit and erecting it. This work was performed cautiously, with preventive measures and administrative controls to preclude damage to any existing equipment and to ensure relative safety for personnel performing the work. Due to the minor scope. of the maintenance activity and the previously mentioned precautions and administrative controls, this maintenance task does not result in an unreviewed safety question. JAF-SE-81-074, Plant Modification F1-80-015 The modification consisted of adding instrumentation to provide redundant indication of containment pressure and level under accident conditions. Additional reactor vessei pressure indication was added to increase the monitored range and recorders were provided for each pressure and level variable. All instrumentation added is seismic Category I, electrical class 1E, and indicates and records continuously in the control room. New penetrations were added to the torus to implement this modification. The instruments were installed per the recommendation and mandates of the NRC. l l l Page 4 of 22

The instrumentation being added by this modification, (1) Reacto r Vessel Pressure, (2) Containment Pressure, and (3) Containment Water Level are similar to existing instrumentation with the exception of the drywell water level monitors of which none presently exist. The FSAR, appendix A (Technical Specifications) shall be revised to include the new instrumentation to Table 3.2-6 (Surveillance Instrumentation). In addition, the drywell level monitor shall be added to Table 4.2-6 and a description of all added instrumentation shall be incorporated into the FSAR - Section 7. The installation details have been scrutinized to account for construction accidents and installation errors. No single accident such as dropping a piece of conduit on a cable tray during plant operations or shorting out terminals in the 09-3 panel during plant outage could damage more than one safety division in compliance with Safety Design Bases, Section 8, FSAR (Electrical Systems) criteria. The completed installation shall undergo construction checkout and preoperational testing to ensure correct operation. JAF-SE-81-075, Plant Modification F1-81-021 This modification consisted of extending all of the plant's outside perimeter security fence (including gates) to a minimum height of eleven (11) feet. This height includes one (1) foot for three (3) strands of four barbed wire. This modification now provides a security fence conforming to the specifications of 10CFR73.2. The work area was restored to full security effectiveness at the close of each work day. The work performed was non-safety related and the activity was out of doors away from the plant. JAF-SE-81-076, Modification M-76/13023) This minor modification consisted of extending the sprinkler system of the warehouse by six (6) heads to complete area sprinkler protection. The system now complies with insurance regulations. This modification will enhance the fire protection of the Fire Protection System and does not alter any part of the FS AR. JAF-SE-81-091, Modification M-12/15112 This minor modification replaced the Reactor Water Cleanup System effluent flow control valves (12-4F CV-15 A & B) with new valves to make maintenance possible. The old valves manufactured by Black, Sivalls & Brysons (BS&B) are out of production making spare parts unavailable. The new Masonellan valves are compatible with other plant flow control valves. Page 5 of 22 )

Th e maintainability of the components will be improved thereby decreasing the probability of component; malfunction. By changing the size of the valve, the flow control' system will operate in a more stable fashion, thus, increasing system reliability'and operability. JAF-SE-82-004, Modification M-16, 27/14930 This minor modification consisted of removing' existing rigid instrumentation conduits, for 16RT D-114 ~ and 27RT D-10 2 A1 ' and replacing and relocating the items with new flexible conduits, in the torus. Cable No. 1PCPBBX018 was damaged in the removal and -replaced with a new cable. Since the existing cable and instruments were functionally unchanged this work was considered an equipment substitution fo r the conduit and supports.= The rerouting installation was seismic class II. The flexibility of the conduit will result' in the seismic load being transferred to the supports. JAF-SE-82-007, Modification M-71/15189 This minor modification consisted. of repairing the EC-1 Series Trip Device due to a malfunction of that device. This work was performed due to the lack of exact r.eplacement parts and plant conditions required that breaker 11508 be placed back-in service at the earliest possible date. J A F-SE-82-008, Portable Demineralizer Temporary Modification The work associated with this safety evaluation involved installing a high flow rate portable demineralizer to enable rapid production of demineralized water _for filling the torus following the refueling. The plant was in the cold shutdown mode for this operation. Since all portions of this installation were temporary and all equipment was returned to normal after completion of water makeup, this installation had no effect on plant safety after it was removed. This temporary demineralizer affected plant systems only'during installation, operation and removal. The following provisions were made due to concerns relative to plant safety. 1. Provisions were made to ensure that demineralizer effluent is of quality appropriate to torus and reactor water 7 2. Provisions were made to ensure that the condensate storage tanks were not drained to levels lower than those required by the FSAR in the event of a temporary line break during makeup operations. Page 6 of 22

3. Provisions were made to ensure that the condensate storage tanks were not overflowed during makeup operations, which could cause contaminated materials to pass to the site. 4. Since a temporary line was run from the condensate transfer line within the turbine building to the screenhouse (an uncontrolled area), precautions were taken to ensure that a temporary line break within the screenhouse would not allow contaminated materials to drain from the CST's back to the screenhouse. JAF-SE-82-011, Modification M-71/11260 This minor modification consisted of installing an rf voltage probe around the main generator neutral at the neutral t ransformer termination cabinet. The probe was then connected during operation to an rf meter in order to record the rf noise produced by the generator. The rf noise is known to be an indicator of incipient fault condition of generator winding failure. The rf monitor of generator has been placed in service at the Authority's Astoria #6 Steam Station and has provided valuable information on the state of the generator. Similar work is proceeding at the Indian Point #3 Unit. Neither installation has caused a degradation in the generator unit protection or insulation levels. A review of these requirements at JAF indicates no degradation in these levels. JAF-SE-82-012, Preoperational Test 17H This test for the High Range Ef fluent Monitoring System encompassed the functional testing of the Victoreen 845 area monitors installed under Modification F1-80-014. Included was the functional testing of the annunciator windows, computer digital alarms, pen recorders and computer analog printouts. The High Range Ef fluent Monitors (HREM) are connected to the low range effluent monitors. At each location the outlet tubing for the low range monitor is tapped into, diverting the sample e f fluent through the HREM. As this is the only connection between the two, testing of the HREM will not affect the low range monitors. The HREM receives power from the 71ACUPS-2 panel. Operation of the circuit breakers being used were verified by overcurrent trip testing prior to installation, therefore; testing of the HREM did not affect this power supply. Conduct of preoperational test no. 17H did not impact any statement in the FS AR nor any technical specification. This test can be performed at any time. Page 2 of 22

JAF-SE-82-015, Modification M-31/9683 This.-minor modification consisted of removing ~ valve 31A0V-V4B and capping the resultant pipes. The valve, associated with the Turbine Moisture Separator Reheaters, is no longer used and was a potential source of steam leakage. JAF-SE-82-017, Prcoperational Test 05B This test was to verify tripping durations and voltage characteristics of the Reactor Protection System (RPS) bus when the MG Sets'are shutdown. This test also verified a major load transient on the RPS bus when its fed from its alternate source. Since the plant was in the cold shutdown position, each power. supply to the RPS was temporarily disconnected for testing as the RPS is not required during cold shutdown. -JAF-SE-82-018, Preoperational Test 27G This test encompassed the functional testing of the Post Accident Sampling System fo r valves that have been irestalled in-line in the Containment Atmosphere Dilution (CAD) System sample tubing. Included were all power supplies and circuit components required for the operation of these valves. This test involved operating the valves. The valves were installed in-line in the oxygen analysis sample tubing of the CAD system. They open when energized, and they must remain open for the' oxygen analyzers to operate. The oxygen analyzers must be operational before startup of the plant; therefore, this test was performed prior to startup of the plant. This test did not impact any statement in the'FSAR nor any Technical Specification. I JAF-SE-82-020, Preoperational Test 27H l This test encompassed the functional testing of the General l Electric Post Accident Pressure.and _ Level Instrumentation [ consisting of the following parameters: 1. Reactor Vessel Pressure 2. Primary Containment Pressure 3. Primary Containment Water Level 4. Suppression Pool Water Level l The test involved simulation of all transmitter input to verify specific signal outputs to the control room indicators and recorders and correct operation. l Page 8 of 22 l l l l

This test did not impact the safety of the plant nor any statement in the FSAR or Technical Specifications. JAF-SE-82-025, Modification M-99/9650 This minor modification consisted of removing the window between the control room viewing gallery and the shift supervisor's office. A small counter top was installed in the opening. This walk up window eliminates a great deal of noise and confusion in the control room and-provides a more appropriate atmosphere for conducting power plant operation. This philosophy is consistent with the Three Mile Island Lessons Learned and has been recommended by both the plant operators and the Human Factors Survey Team. The FSAR and Technical Specifications do not address the security system. The wall separating the viewing gallery and the control room is not a required security barrier. The wall lies within the physical barrier of the control room required by 10CFR73 - Physical Protection of Plants and Materials - and therefore serves no purpose with respect t o Security. JAF-SE-82-026, Modification M-10/16174 & M-10/14833 This minor modification consisted of substituting the original component solenoid for 10A0V-68A & B. The original component was an ASCO solenoid Cal. No. WPHT8300 B64-G. The replacement solenoid is an ASCO NP8320A184E. The function of the disk operation is only to verify the free movement of the LPCI testable check valve shaft. The solenoid does not perform any safety related function or in any way affect the pressure boundary or safety related function of the LPCI testable check valves 10A0V-68A &B. JAF-SE-82-27, Hodification M-05/13025 This safety evaluation addressed several questions raised by the NRC concerning repair work previously done under WR #05/13025 (see Safety Evaluation JAF-SE-82-033 - Letter to R. C. Haynes l from C. A. McNeill, Jr., dated 5/6/82). The following items are discussed and will supplement the referenced safety evaluation, a) What qualifications are applicable to the splices used in the repair of Cable 1RPSCUC1587 b) Will periodic testing be performed by NYPA personnel on the subject repairs or cable? I Page 9 of 22 l l

c)- What-is the failure mode of the cable / splice (e.g., hot shorts, ground f aults or loss 'of continuity)? And what means are available for failure detection? d) What is the reasoning for not replacing the cable as presently in~ stalled? Basis for Evaluation: a)- The splice repair performed on cable 1RPSCUC158 was done in accordance with Maintenanc e Procedure MP-71.4, " Installation of Press,uSe Terminal Electrical Connectors". This procedure references and uses as a basis the Thomas & Betts Specification No. -QPS-TB (CH)-878, Rev.-0, " Qualification Program Specification for T&B Pressure Terminal Connectors, Tefzel Insulated STA-Kon Terminals, for Use in Class 1E Applications within a Nuclear Power Generating Station". Sect ~ ion 3.4 of the T&B Specification' states that the qualified life, as defined in IEEE 323, is 40 years of continuous operation. Additionally _the method of applying insulation tape was performed in accordance with Maintenance Procedure MP 71.1, " Termination-& Insulation of Electrical Power & Control Cables". The methods described in MP-71.1 are qualified to IEEE-323. b) Insomuch as the failure mode of the cable repair, cable and system are all fail safe (i.e., a half scram condition shall result from a ground fault or.a loss of continuity only on the Group 3 Control Rods) it is.the conclusion that no periodic continuity or insulation resistance checks need be performed on Cable #1RPSCUC158. c) The failure mode of the repaired cable and the reactor protection system associated with this cable is fail safe. In the unlikely event of a ground fault on the repaired cable (1RPSCUC158), the result will be a half scram condition on the ' group 3 scram solenoids. This condition is readily detected from control indicators. d) The basis'for not proceeding with the replacement of cable 1RPSCUC158 is due to the difficulty in removing the presently installed cable. The difficulty arises from the installation of a new fire penetration sealing' material. The new material is ~ relatively' dense and must be partially removed from the wall sleeves that are used to route'the subject cable. Because of the di f ficulty in removal of-the fire seal, it is nearly impossible to remove the repaired cable. For this reason and the other reasons stated above, we conclude that replacement of cable 1RPSCUC158 is unnecessary and will continue to perform its intended safety function. Page 19 of 22

f JAF-{ 92-028' This safety evaluation was written for.a revision to a procedure. ihe change involved a revision-to the radiation protection i' procedure to raise contamination limits for unrestricted releases and the plant from 100dpm/100 cm2, removable and less than twice the background count rate (background-less than 300 cpm) to 1000 i dpm/100 cm2 removable and 5000 dpm/100 cm2 fixed and' removable.. This change is based on recommendations-of the United States Nuclear Regulatory-Commission that facility contamination limits for unrestricted release be set not_less than 1,000 dpm/100 cm2 ~ for removable surface contamination and 5,000 dpm/100 cm2 total-i (fixed +-removable). 'This change will'also result'in a reduction of the amount of material disposed of as solid low-4 level radioactive waste.. JAF-SE-82-029, Modification M-66/13316 This minor modification consisted of upgrading the insulation class for the Crescent Area unit cooler drive motors (66UC-22A thru K) from class F to class H to improve qualification margin with respect to accident ambient temperature conditions. These unit coolers provide cooling in areas containing essential ECCS l components required for long tern post accident operation. i Upgrading the insulation class will permit the-motors to operate 4 at the higher ambient temperatures calculated to occur following postulated high energy line break accidents in the Reactor Building. It does not affect in any other way the performance of these motors. JAF-SE-82-031, Preoperational Test 16E This safety evaluation is for the test used to gather data which was utilized to determine pressure loading and strain response of the torus'shell during relief valve actuation. I This evaluation covers the lifting of a relief valve during power operations. Evaluation of the test equipment installation and i modification of the torus and safety relief discharge lines has been previously performed as part of the plant modification (F1-81-024). The limiting failure to occur as a direct result of this testing is a stuck open relief valve.- This is an abnormal operational transient involving decreasing coolant inventory analyzed in the FSAR, Section 14 and subsequent reload analyses. The results of these analyses are acceptable. Page JJ of 22

E Operation in accordance with the Special Safety Precautions section of this test assures the torus is. maintained as a heat sink and source of. emergency makeup water as prescribed.in the . Technical Specifications Limiting Conditions for Operation for Containment Systems. JAF-SE-82-032, Modification M-12/13365 This minor modification consisted of repiping sample flow control valve and flow-indicator from the sample rack 25-3 2 to the outlet of the constant temperature bath. This change added reliability to the conductivity cells which perform no automatic functions. Relocation of the flow control valve' upstream of the conductivity cell prevents overpressure without changing the function of the system. JAF-SE-82-035, Modification M-69/15981 This minor modification consisted of installing temporary radiation shielding and associated supports of radwaste tank ventilation line 46'-6"-VR-136-10 in the Radwaste Building Control Room to reduce personnel radiation exposure. A permanent modification consisting of replacing and rerouting the subject pipe along with replacing other bondstrand pipe is being planned for the future. For the highly unlikely condition of a postulated seismic event, or other postulated event, which could possibly cause piping system failure, there is no risk to public health and safety, as no Safety Class 1 equipment is located near, or under the subject piping. This piping is not required to be Seismic I (JAFNPP FSAR Sections 12.2.3 and 12.2.5) nor is it needed for the safe shutdown of the plant. 1 1. JAF-SE-82-036, Plant Modification F1-82-013 This modification consisted of replacing the three existing Control Room access doors with new bullet resistant doors per 10CFR73.2. The design of the door installation took into account the seismic loads imposed under OBE and DBE conditions. JAF-SE-82-040, Test This test involved connecting a radio noise meter to a radio frequency ( r f) voltage probe already in place on the neutral of the plant generator. Measurements were taken of the radio noise produced by the generator in the frequency range of 10KHz to 12MHz. A second element of the test had the radio noise of the generator monitored for a one week period by using a temporary strip chart recorder. This test assisted in the determination of the insulation integrity of the generator. Page 12 of 22 v y - -w,- 1,i, .v. ,r-r,v-ir ,,w-w,,w-n-. ,. - +, .-y w e y e-wy, -w-,

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The connection of the radio noise meter to the rf voltage probe already installed on the neutral of the generator will not cause a system protection degradation. Even if the output of the voltage probe were grounded there would be no effect on the differential relay circuit of the unit. There were no nuclear safety systems effected by this testing. JAF-SE-82-041, Modifications M-34/16218 & M-34/16219 This minor modification consisted of adding Furmanite Nuclear Grade F-100 and/or X-1 compound t o 34NRV-111 A & B hinge pin cover gasket areas. Temporary Maintenance Procedure T-19 was used to per form the repairs. Past experience with injecting Furmanite compounds in stuffing boxes of valves has shown that Furmanite compounds will not cause a valve stem to bind when operated. Furmanite Corporation of America hos reviewed the design of the valves and concluded that the hinge pin covers and valve body can be drilled per their procedures without degrading the structural integrity of the valves. The Furmanite compounds will not prevent the valve disc from closing and mating with the valve seat. This work enhances the reliability of the valves to perform as designed. JAF-SE-82-042, Modification M-00/13627 In supporting the NRCB 79-01B Equipment Qualification Program at JAF, seventeen (17) Hoffman boxes were modified to house a seven day temperature recorder ( b at t e ry operated) and a radiation monitor (TLD). These JB's were mounted throughout the plant to monitor temperature and radiation levels. The mounting of the JB's was in accordance with seismic class II requirements. The proceeding evaluation is based on a review of t;.e JAF FSAR for class I structures and seismic design, section s 12.2.2, 12.4.6.1 and 12.5.1. This minor modification did not alter the F S AR as written and did not create an unreviewed safety question pursuant to 10CF R5 0.5 9. JAF-SE-8?-043, Plant Modification F1-81-016 This modification consisted of installing a multitude of eight-hour battery packs with 2-twelve watt halogen lamps to obtain lighting levels of 1/2 to 1 foot candle for access to equipment and approximately 3 foot candle for equipment operation. Page 13 of 22 l

l For' operation of equipment inside the drywell; 2 portable 1 rechargeable lanterns will be housed in a cabinet outside the drywell personnel hatch.- In addition, 2 portable lanterns will be maintained in the control room for operator use in the general-plant. Both these' cabinets will be fire-equipment red with a breakglass : front access. The overall lighting design was developed to comply with 10CFR50 Appendix R. The battery packs and portable lanterns will be supplied from the 120V AC lighting circuits in the area, so that the lights will automatically go on should AC lighting in the area fail. In general, the Emergency Lighting Battery. Pack System is not required to be a seismic installation. Only due to the weight of the battery packs; in areas where there is safety related i equipment, battery packs were mounted on seismic supports. Wiring and raceway.were installed in accordance with original design by Stone & Webster. JAF-SE-82-044, Modification M-23/16793 s! This safety evaluation contains the basis for changing the i setpoint on the HPCI switches 23DPIS76/77 from <282 inches of H O to <106 inches of H 0. An error in the setpoint 2 2 calculation was not discovered in the administrative review during the startup test program. The HPCI test' program conducted at the plant from March 12-17 has H O will be i verified that the trip setpoint of 106'in. 2 conservative enough to trip the turbine on excessive steam flow from a HELB. The setting also meets the requirements of the technical specifications'to trip the turbine at a steam flow of less than or equal to 300% of steady state steam flow. JAF-SE-82-045, Operability Test 70A i This safety evaluation encompasses a special stroke testing of 70 MOD-113. The test requires the control room ventilation system to operate in the " Isolate" mode which will pressurize the control room and force MOD-113 and MOD-114 to modulate. The minimum opening of the outside air damper MOD-113 will be noted and recorded. i Existing site Technical Specifications, NRC Regulatory Guides and the Final Safety Analysis Report were researched in order to i determine whether or not this test would affect any system design bases, increase the possibility of safety hazards or increase th e j possibility of accidents considered in these documents. .i Page 14 of 22

During the short interval when MOD-113 was forced closed,'it would not'be possible to. pressurize the control room. Sincoithis condition existed for a maximum o f 15 seconds, this time duration is less than the closure time on isolation dampers MOV-107 and MOV-108. There for e, during the test, the control. room could be re-pressurized faster than the normal isolation time. JAF-SE-82-046, Preoperational Test 01-107A This test was to demonstrate the proper functioning of the Recombiner and Of f-Gas System Pressure Reduction Modification components,. controls.and circuitry before placing the= system into permanent service. Verification of most operating parameters-were conducted both in the warmup and normal system operating modes. No assumption used in the FSAR, Section 11.4 - Gaseous Radioactive Waste System specifically 11.4.4.2,: relating to condenser off gas system were changed. The safety margins defined in the Bases of the Technical' Specifications were not reduced and the results of an accident analysis were-nachanged.- JAF-SE-82-047, Modification M-99/11228 This minor modification consisted of providing a security -intrusion alarm on roll-up doo'r R-300-6; located in MG. Set Room on the East wall. The addition of this intrusion alarm was necessary to fulfill the requirements of 10CFR73.5 5 and is consistent with security procedure 5.1.5. The modification design and installation sequence have been reviewed against the FSAR and other applicable documents and it was concluded that there was no effect whatsoever on existing safety and design analyses. JAF-SE-82-048, Modification M-17/11335 This minor modification consisted of installing an interim dose assessment system using meterological _ data acquired from Nine Mile Point I and JAF meterological towers. The interim system included telephone access to an off-site computer for calculation of' release dose rate' assessment. This modification replaced the existing recorder in the Control Room EMRP Panel with new recorders and a demultiplexer. A computer terminal and video screen for computer access is installed near the EMRP panel and j in the Technical Support Center. 5 i Implementation of this minor modification does not constitute an unreviewed safety question pursuant to 10CFR50.59. i l Page 15 of 22 [

~. c j ~3AF-SE-82-049, Preoperational Test 72A-This test for the Technical Support' Center Filtered: Ventilation System encompassed theofunctional testing of filter unit 72-F-28 'and its associated components which were installed in order to . satisfy the requirements of NUREG-0696, Section 2.6. Included was a verification.of=the adequacy of the Technical Support Center pressure boundary'and.a verification that the flow rate 4 'under both clean and dirty filter. conditions is within +20% (as pe r ANSI-N510-197 5) of_the design ~ rate of'300 CFM. This i preoperational test also included efficiency-testing:of!the charcoal and HEPA. filters. ~ l 'This preoperational-test will affect only the-Administration. Building Ventilation and Cooling System (No. 72). Conduct of the; test will not impact any statement in the FSAR nor any Technical Specification. JAF-SE-82-050, Preoperational Test 27F This test was conducted to functionally prove the design-and construction ~ of modifications oto the Primary - Containment Purge Valve System and the controlling circuitry. The test was conducted during plant shutdown and demonstrated the isolation logic for closing the valves and the individual signal overrides and verify the response of these valves. The Nuclear Safety Evaluations for Modifications 80-28, 81-19, 81-49 and 80-16 considered the affects of the modifications on the overall safety of the plant. Essentially these modifications increase the margin of safety. As stated by the design objectives for the Primary Containment and Reactor Vessel Isolation Control System FSAR, Section 7.3.1 and fulfill the' design bases (FSAR Sections 5.2, 7.2, 7.3, and 8). The preoperational test is the functional proving of the completed i modifications and is the final check on'the installation before placing the modified valves and control circuits into service, thus fulfillingLthe' inspection and testing criteria for the i PCRVICS (FSAR, Section 7.3.6). The preoperational test was performed during shutdown and no single malfunction such as shorting out wires in a control l circuit would affect any more than one safety system. I JAF-SE-82-052 This maintenance activity involved the removal and jumpering of two LPCI independent power supply battery cells that cannot meet minimum operating characteristic specifications. i Page 16 o f 22 Il i l.. ..a_,._.,.

E -i The 'two subject cells are _ par'. of. the LPCI Battery and 'are not capable of meeting the minimum acceptance: specification for individual cell voltage (ICV) o f 2.12. volts per - cel-1. 0ther corrective maintenance options have failed to remedy this-low ICV- ~ condition.- The maintenance activity serves as'an interim measure to. allow removal of-the-out of specification cells. prior.to-receiving new. replacement-cells. Removal ~(by jumpering out) two of the LPCI battery cells will.not decrease'the discharge rate of 78 amperes based on the regulating capability of the' inverter. It slightly decreases the overall power available,.but by less than a factor of.1/93. In addition the capacity of the LPCI battery system is sized much larger'than required by.its loads. Therefore, based on the above evaluation, the' removal of two cells the battery will still be able to perform its intended safety objective as stated in the FSAR. JAF-SE-82-053 This safety analysis covers the analysis performed to determine the. feasibility of simultaneous two-pump operation of the Standby Liquid Control System with the present system configuration. General Electric performed an analysis on this subject. This safety analysis was written to put the results in.the. FitzPatrick's safety evaluation format and therefore administratively ensure that 10CFR50.59 requirements were met. The intent of the requirement is to decrease the time in which the neutron absorbing solution is injected into the reactor. However, an upper limit of 20 ppm / min. change in boron concentration in the reactor and recirculation loop rater is specified in the SLCS Design Specification 22A2896, Revision 1, Page 4). This corresponds with a 50 minute minimum injection time and a maximum. injection rate of 64.4 gpm. The increased flow rate resultant from two pump operation would exceed this maximum injection rate. A re-analysis'would have to be performed 'to ensure mixing in the reactor would be sufficient at this new fl ow rate of approximately 106 gpm. Calculations concluded that the available NPSH, with the present design, is inadequate for two-pump operation. Unless a desireable phase angle can be maintained between the two triplex pumps the acceleration head loss would be too large for simultaneous operation. GE's analysis also expressed a concern for the increased pressure drop in the discharge piping which would be resultant from the increase flow. However, the calculations indicated that this increase would not prohibit two-pump operation. Page il of 22

In conclusion, simultaneous two-pump operation was not recommended unless these concerns (solution mixing and HPSH) are addressed in detailed re-evaluation and modifications. To implement two pump operation at this time would constitute an unreviewed safety question. JAF-SE-82-054, Modification M1-82-059 This minor modification provided a means for obtaining samples for analysis at the suction of the potable water pump 74-P-4. This modification was to a low-pressure non-safety related system. The connection will be used to periodically sample the potable water tank (74-TK-2) to verify water quality. Piping is non-seismic, clas s Q3. JAF-SE-82-057, Modification M1-82-074 This minor modification consisted of raising the setpoint of 01-107-PS-100 from 25 psig to 50 psig. This allowed the recombiner system pressure to be raised to 45 psig to aid in the proper regeneration of the off-gas dryer dessicant. This is a temporary modification required to assess system performance at a higher back-pressure. JAF-SE-82-058, Modification M1-82-082 This minor modification consisted of drilling 0.25" diameter weep holes at the low point in various junction and terminal boxes in the Primary Containment and Steam Tunnel. This would prevent the possibility of box implosion and moisture accumulation during and after postulated line break accidents. This 'todification will improve the safety performance of the associated equipment by upgrading the boxes to a configuration which has been type-tested for LOCA environments per IEEE Standard 323., f Page 18 of 22

J A F-SE-82-06 0, Modification M1-82-080 This minor modification remedied the unreliability problem of the actuation of the MSIV 90% RPS position switches by performing the following: 1) The existing maintained contact snap lock switch was replaced with a similar switch which is only actuated by arm movement in the clockwise direction. The switch arm is equipped with'a roller ass emb ly.- 2) The existing striker was modified by the installation of a ramp as s emb ly. The ramp was fabricated of steel plate and bolted to the striker plate. The switch will be actuated by movement of the roller up the new ramp. 3) The existing switch seal assembly was replaced. This minor hardware change resulting from the switch replacement and method of actuation does not affect the FSAR in any way. The plant technical specifications were reviewed and no additions or deletions are required as a result of this minor modification. All components are qualified for this specific class 1E application. All replacement components are environmentally qualified and seismically qualified for this application. JAF-SE-82-061, Modification M1-82-081 This minor modification involved the addition of an extra stem disk pin to the "D" inboard main steam isolation valve. Maintenance Procedure MP 1.3 was revised to incorporate this modification. The revision provides.a more specific method for pin installation in all the main steam isolation valves. No structural effects were placed on the stem. The additional pin improves the anti-rotation effect. There was no change in the safety analysis due to this modification. JAF-SE-82-062, Preoperational Test 27E This test was designed to verify that the high and high-high trips, for the Post Accident Contain;nent High Range Radiation Monitoring System, provide the appropriate annunciation and actuation signals fo r isolation test conditions involved actual isolation functions during plant shutdown. The sections of the FSAR af fected by the preoperational test are included in Section 5 under " Containment". Since this system provides another trip signal for existing isolation functions, no non-conservative changes have occurred. Page 19 of 22 i

r JAF-SE-82-064, Modification M1-82-083 This minor modification remedied the installed connections in relay 27L which did not permit correct operation of computer point D028, which indicates the condition of the 115KV line to Lighthouse Hill. The modification removed the jumper from terminals #6 and #8. Removed connection from terminal #4 and connected it to terminal #8. Removed connection from terminal #1 and connected it to terminal #10. This modification did not cause any trips or closures and its sole purpose is to indicate a live or dead line to the computer.. This modification was performed on a non-safety related system. JAF-SE-82-070, Modification M1-82-093 This minor modification involved the substitution of 9 grafoil gasket for the malleable iron bonnet gasket of HPCI valve 23-A0V-53. The critical characteristics of valve 23-A0V-53 are it's differential pressure sizing, maximum pressure, normal and maximum temperature, and environmental compatibilities. The mechanical properties of grafoil exceed the critical requirements of 23-A0V-53 in radiation atmospheres. Therefore, the use of grafoil as.a substitute for the malleable iron bonnet gasket provides an equal to or better bonnet seal-than the malleable iron seal. JAF-SE-82-071, Modification M1-82-094 This minor modification involved the addition of an extra disk lock pin to the "D" inboard main steam isolation valve. Maintenance Procedure MP 1.3 was revised to incorporate this modification. This revision provides a more specific method fo r pin installation in all the mainsteam isolation valves. No significant structural effect is placed on the piston assembly. The additional pin provides anti-rotational effects and there is no change in the safety analysis. Modification M-20/17313 This minor modification consisted of replacing the 2 inch suction line ( 2"-SY-12 3-19 6 ) from the bottom tank flange of the spent Resin Tank (TK-76) to the suction valve (A0V-315) due to a hole in the line below the flange. The work was performed on a non-safety related system. Page 20 of 22

e Modification M-41/11415 This minor modification consisted of removing the Y-strainers in various lines and replacing them with piping of the same class. The Y-strainers were a continuous source of steam leakage. The work performed was on a non-safety related system. Modification M-46/17197 This minor modification consisted of replacing an existing 6 inch reversible elbow on the service water line (46-6-WS-151-49) with a 6 inch reversible elbow that has 2-2 1/2 inch connections with isolation valves. This will be used to feed the mobile demineralizer for makeup when needed. Th e work was performed on a non-safety related system. Modification M-78/10249 This minor modification consisted of installing an emergency drenching shower and eyewash station in the immediate area of the acid / caustic room on elevation 250'-0" in the Radwaste Building. The water supply is city water from an existing drain valve in the screenwell. The work was performed on a non-safety related system. Plant Modification F1-78-038 This modification consisted of replacing shafts, discharge pipe columns and motor mounts for all three service water pumps. This modification tends to eliminate vibration problems experienced with these pump s. These changes reinforce the structural stability of the pumps. Plant Modification F1-79-007 This modification performed the following changes to the suppression pool (Torus): 1. Installed a structural steel support saddle at each ring l girder. The torus was tied down to the concrete mat with anchor bolt s. 2. Installed a 30" O pipe deflector under the existing vent header. 3. Relocated two safety relief discharge line ramsheads with connecting pipe and realigning a short section of horizontal run of pipe including the ramshead in each of the remaining nine safety relief discharge lines. I Page 21 of 22 i

p.. .These modifications constitute changes as described in the SAR that did not show torus support saddles or vent header deflector. The probability of occurrence or the consequence of an accident evaluated in the SAR but has decreased by the addition of a support saddle and a-vent. header de flector. The SRV line relocations have no effect on the steam condensation capabilities of the pressure relief system as described in the SAR. The possibility of an accident or a. malfunction of a different type than any evaluated previously in the SAR has not been created. The margin of safety as defined in the basis for any Technical Specification has not been reduced. Since those modifications are structural in nature they do not affect any safety-limits nor any limiting conditions for operations. Conclusions above are based on the fact t' hat the only safety considerations affected by this. modification is that the stresses in the torus structures are reduced to within code allowables for the loadings considered. The loads used in the structural analysis are from the General Electric Company's Final Load Definition Report (FLDR) which was prepared as a result of their ef fort on the Mark I Containment Program. The modifications are being performed now in order to meet schedules imposed by the NRC. They have been identified by Teledyne Engineering Services (TE S) as necessary to bring structural stresses resulting from LOCA pool swell to within code allowable values. Continuing review of the FLDR by TES may identify additional requirements for modifications in the structure. Page 22 of 22}}