ML20148N467
| ML20148N467 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/31/1987 |
| From: | Mcdonald R ALABAMA POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8804060526 | |
| Download: ML20148N467 (29) | |
Text
{{#Wiki_filter:. - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ h 8 ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 2 - AMfUAL REPORT REQUIRED BY 10CFR50.59 Section 50.59 of Part 50, Licensing of Production and Utilization Facilities, of the regulations of the United States Nuclear Regulatory Commission, states that the holder of a license authorizing operation of a production or utilization facility may (1) make changes in the facility as described in the safety analysis report, and (2) make changes in the procedures as described in the safety analysis report, and (3) 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 incorporated in the license or an unreviewed safety question (as defined in 10CFR50.59). The licensee is required to maintain records of such changes, tests or experiments, and those records are required to include written safety evaluations which provide the basis for the determination that the changes, tests or experiments do not involve any unreviewed safety questions. Brief descriptions and a summary of the safety evaluations of the changes, tests or experiments as described above, for the Joseph M. Farley Nuclear Plant Unit 2 which were completed in 1987, are provided in the following. i l REHidst-D1.45 1 I I 1 8804060526 31 Ll PDR ADDCK O O 364 R PDR g
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Subject:
PCR/PCN 84-2609 (S84-0-2609)
Description:
Upgraded the existing Primary Meteorological Tower instrumentation. his included the addition of redundant wind speed arxl wind direction channels at the 150 foot elevation. Safety Evaluation his modification increases the reliability of the Primary Meteorological Tower instrumentation. Also, the addition of redundant wind speed and , wind direction instruments prevents entering a Technical Specification Limitin Condition for Operation upon failure of a sin le channel. PORC Review PORC Meeting 1522, 5/13/86
Subject:
PCR/PCN 84-2865 (B84-2-2865)
Description:
Installed a three-inch stainless steel valve Detween the containment sump pumps and the containment isolation valve (HV3376). Safety Evaluation: he addition of this valve eliminates the need for personnel to enter the sump area for local leak rate testing or to isolate the sump pung discharge valves. his will reduce the radiation exposure to plant personnel. PORC Review PORC Meeting 1647, 3/12/87
Subject:
PCR/PCN 84-2895 (S84-2-2895)
Description:
Replaced the existing air compressor temperature switches with thermocouple sensors and remote temperature controllers. It is believed that some air compressor trips have resulted from setpoint changes on the current instruments. %ese setpoint changes are believed to be induced by piping vibration. Safety Evaluation his modification v".11 eliminate setpoint changes by instrument vibration and provide more reliable air tenperature control of the air compressors. PORC Review PORC Meeting 1493, 2/25/86
Subject:
PCR/PCN 84-2915 (B84-2-2915)
Description:
Installed isolation valves on the service water supply and return piping for the following :oom coolers: RHR pumps 2A and 28, battery chargers, and MCC 2A and 2B. Safety Evaluation %ese valves will facilitate cleaning and/or replacement of the service water piping. %ese valves will have a negligible effect on system performance. Precautions will be taken to ensure that these valves remain open when the associated lines are in service. PORC Review PORC Meeting 1667, 4/28/87
Subject:
PCR/PCN 86-3496 (S86-0-3496) nescription: Erected a solidification and dewatering facility (SDP). Safety Evaluation: 2e SDr provides the capability to transfer primary resin, steam generator blowdown resin, chemical drains and evaporator bottoms to four bulk disposal liners for solidification or dewatering. No new liquid or gaseous effluent pathways were created. %e four disposal liners are located in shielded pits which are vented to the Unit 1 Auxiliary Building Radwaste Area Ventilation System. W e building sump is pumped to the Unit 1 Auxiliary Building Floor Drain System. We SDF also provides for interim storage of up to 1690 gallons of contaminated oil. We operation of this facility does not affect nuclear safety. PORC Review PORC Meeting 1554, 8/8/86 PORC Meeting 1627, 1/23/87 PORC Meeting 1616, 3/10/87 PORC Meeting 1663, 4/20/87 subject: PCVPoi 86-3499 (B86-2-3499)
Description:
Provided for the automatic sequencing of the swing battery charger (2C) to the diesel generators in ' case of a loss of offsite power (14SP), t
safety Evaluation: his modification eliminated operator action during plant shutdown following an LOSP if battery charger 2C is in operation. %e loading of the diesel generator is not affected by this modification since swing battery charger 2C operates only in place of and never in addition to train dedicated battery chargers (2A and 28). PORC Review PORC Meeting 1673, 5/14/87
Subject:
PCR/ Poi 86-3523 (886-0-3523)
Description:
Removed all the existing equipment in the decontamination room and installed a high pressure water spray decontamination unit. Items included in this unit are a stainless steel sink, several stainless steel shelves, and two receptacles for the spray unit. Also, a vent valve was added on the demineralized water system supply piping to the high pressure water spray decontamination system. Safety Evaluation: his new system replaced an out-moded decontamination system. %e addition of a vent valve will provide the ability to vent the piping during a hydrostatic test or during initial operation of the decontamination system. his new water spray unit will enhance equipment decontamination and decrease the dependency on respiratory protection by decontamination workers. PORC Review: PORC Meeting 1611, 12/4/86 PORC Meeting 1624, 1/15/87 PORC Meeting 1645, 3/5/87
Subject:
PCR/Pai 86-3635 (S86-0-3635)
Description:
Replaced an auxiliary relay in the control circuit of well water pump #3 with a time delay to prevent tripping the pump from momentary surges in header pressure. Safety Evaluation: his modification did not involve any safety related equipnent. his will improve the reliability of well water p;mp #3. PORC Review: PORC Meeting 1606, 11/25/86 I [
Subject:
PCR/PCN 86-3671 (S86-2-3671)
Description:
Replaced the existing carbon steel portion of the condensate header sample line to the chemistry laboratory with stainless steel tubing. Also, installed local sample points for the pump suction, pump discharge and the combined discharge header. Safety Evaluation: he modifications performed will enhance plant operation by providing accurate condensate water analysis. PORC Review PORC Meeting 1595, 11/11/86
Subject:
PCP/PCN 86-3695 IB86-2-3695)
Description:
Upgraded the existing capacity of the new fuel monorail hoist from 2000 pounds to 2500 pounds. his modification involved the installation of a new motor, load block, hoist cable, and capacity plates. Safety Evaluation: mis upgrade is needed because FNP plans to routinely lift loads greater than 2000 pounds. A structural evaluation was performed to determine whether the increased lift capacity would impart some uplift load on the new fuel storage racks. It has been concluded that the new fuel storage racks are capable of withstanding uplift loads in excess of the new hoist capacity of 2500 pounds without failure. PORC Review: PORC Meeting 1715, 8/13/87
Subject:
PCR/PCN 86-3791 (B86-2-3791) 1
Description:
Added 18 feet of ductwork tn provide cooling to the Unit 2 Hot Shutdown Panel /Comnunication Room (2202). Bis ductwork is connected to the existing Unit 2 Conputer Room (2201) ductwork. 1 4 1 _
Safety Evaluation: The additional cooling to the hot shutdown panels will reduce the failure rate of the electronic components. We computer toon heating and ventilation air conditioning systems are not engineered safeguard system 6. No credit is taken for their operation in analyzing the consequences of an accident and failure of these systems will not affect the plant's safe shutdowr capability. Additionally, the hot shutdown panels can perform their safety fvnetions without air conditioning. PORC Review PORC Meeting 1667, 4/28/87
Subject:
PCR/PCN 86-3793 (B06- 2-3793)
Description:
Modified the post accident sampling system (PASS) to make its operation sim lar to Unit 1. i his modification involved the replacement of the Hoke valves with riupro bellows rialed valves and the addition of flow indication to the system. Safety Evaluation he modification improves the operation of the PASS. PORC Review PORC Meeting 1675, 5/19/87 PORC Meeting 1720, 8/25/87 PORC Meeting 1728, 9/15/b7
Subject:
PCR/PCN 86-3923 (B86-2-3923)
Description:
Replaced the existing containment tempeinture monitor TISH-3192 and recorder n-3188 with a new digital recorder. Safety Evaluations his instrumentation will provide more versatility and convenience to plant operators in recording containment temperatures. PORC Review PORC Meeting 1658, 4/9/87
subject: POR/PCN 86-3930, (S86-2-3930)
Description:
Installed carpet in the Unit 2 portion of the main control room at-the-controls area. %e elevated portions which constitute the operator 1 and shift supervisor work stations are carpeted as well. Safety Evaluation: We carpet will reduce operator fatigue and reduce the noise level in the control room. he addition of the carpet will not significantly affect the fire safety analysis of the control room. Wis design change will not interfere with or degrade other systems. PORC Review PORC Meeting 1667, 4/28/87
Subject:
PCR/PCN 87-4228 (B87-2-4228)
Description:
Replaced the existing bolts on all three steam generator primary manway covers with a stud and nut arrangement. Safety Evaluation he replacement studs and nuts are sized to require no modification of the covers, gaskets, or the threaded holes in the steam generator. %e design, fabrication, materials, and inspection of the studs and nuts meet ASME Boiler and Pressure vessel Code requirements as specified in the reference drawings. We loads on the studs, nuts, washers, covers, gaskets, and the steam generator have been evaluated as acceptable. We stresses , due to the preload on the studs and design loads of the steam generator are determined to be less than the allowable limits as specified in Section III of the ASME Code. Also, studs of a similar design have been assessed to have a fatigue ).ife equal to or greater than the fatigue life of the original bolts. PORC Review PORC Meeting 1762, 11/5/87 I
Subject:
PCR/PCN 87-4272 (B87-2-4272)
Description:
Replaced the existing manipulator crane Dillon load control system with a computer control hi load control systein. We new system has data collection capabilities and improved control for reducing fuel assembly grid strap danage during fuel movement. Safety Evaluation: h is system allows monitoring and controlling of actual loads, plua providing a historical record of the loading on fual handled by the machine. It automatically coupensates for the constantly changing resistance to mo e.:.ent as the fuel assembly is raised or lowered, keeping the over or underload trip limit on the fuel at a constant value. PORC Review PORC Meeting 1738, 10/1/07
Subject:
PCR/PCN 87-4352, (B87-2-4352)
Description:
Installed Teledyne big-beam 8-hour emergency lights to provide light for manual operatic" of MOV3094A-B and MOV3094B-A (spent fuel p l cooling to the opposite train of component coo ing water) located in rooms 2445 and 2422. Also, lights were installed to facilitate operator access to these rooms. Safety Evaluation: he installation of these 8-hour emergency lights meets the requirement of 10CrR50 Appendix R. PORC Review: PORC Heating 1699, 7/10/87 Scbject: PCP/PCN 87-4384 (B87-0-4384) i
Description:
Replaced the existing comercial grade Anastat relay Model 7012PA with a seismically qualified Agastat relay Model E7012PA002. he new relay is installed in the Unit 2 600V errorgency inad center 2E and is part of the overcurrent protei n.on of the preferred power supply to Unit 2 - Train B 600V emergency load center 2E. It performs a safety related function. However, for tinit 1, this relay is part of the overcurrent protection of the alternate power supply to the normal section of 600V load center lA, and it does not perform a safety related function. ,
)
Safety Evaluation: he new relay E7012PA002 is seismically qualified by testing and is documented in test reports. 2ese test reports show that the new relay is qualified to higher accelerations than required for all modes of relay operation. In addition, the new relay installed on Unit 2 does not affect the load center's original seismic qualification since the location, mass, and mounting do not change from a structural standpoint. PORC Review: PORC Meeting 1711, 8/4/87
Subject:
PCR/PCN 87-4491 (B87-2-4491)
Description:
Replaced the flanged spool pieces on the inlets and outlets of the containment coolers with flexible hoses to reduce piping loads on the containment cooler nozzles. Safety Evaluation: The use of flexible netal hose asserblies has been analyzed and it has been determined that all containment air cocler nozzle loads are within the maximum limits provided by the air cooler manufacturer. We related service water piping was analyzed and all piping stresses are within code allowables. PORC Review: PORC Meeting 1754, 10/27/87 l
Subject:
FNP-0-AP-19
Description:
Deleted this administrative procedure which consisted of a. list of procedures that implemented various chapters of the Operations Quality Assurance Policy Manual (OCAPM). Safety Evaluation: EWP-0-AP-19 was developed as an administrative tool. Its purpose was to aid in establishing the conunitments stated in the 00APM in administrative procedures or startup standards. Currently, chc.nges in the OQAPM are implemented through administrative procedures by the manager responsible for the affected procedure. PORC Review: PORC Meeting 1621, 1/9/87
Subject:
FNP-0-AP-76 Revision 4
Description:
Authorized the use of morpholine / boric acid in the secondary water chemistry control system. Safety Evaluation: Morpholine will result in better pH control in the steam and condensate system, which in turn will reduce the amount of corrosion products transported to the steam generators. We joint use of hydrazine, boric acid, and morpholine does not create corrosive conditions for steam generator or fsedtrain materials. PORC Review: PORC Meeting 1760, 11/3/87
Subject:
EWP-2-CCP-325 Revision 2, 'Im 2A
Description:
Provided for installing Chicago fittings with isolation valves on the condansate pump suction strainer drain valves. his allows for boric acid addition to the condensate and feedwater system. Safety Evaluation: We condensate system is not necessary for safe shutdown of the plant. Int tructions are provided for removing the fittings and valves at the completion of the boron addition. PORC Review: PORC Meeting 1624, 1/15/87
Subject:
FWP-0-CCP-708 TCN 2B, TCN 2J, and Toi 3D
Description:
ICNs 2B and 2J provided for the continuous chlorination of the Unit 2 service water system using sodium hypochlorite to reduce the number of Corbicula (clams) in the system. No biocide will be added to the Unit 1 service water system during this treatment period of Unit 2. TCN 3D allowed the addition of chlorine dioxide (biocide) to Unit 1 to be continued during the hypochlorination of Unit 2. Safety Evaluation: Sodium hypochlorite addition can be quickly terminated if deemed necessary. Using sodium hypochlorite eliminates the personnel safety hazards associated with chlorine gas. Emergency actions and precautions already exist in this procedure, he use of both chlorine dioxide and sodium hypochlorite for prevention of fouling in the service water system is already addressed in the FSAR. PORC Review: PORC Meeting 1654, 4/3/87 PORC Meeting 1698, 7/9/87 PORC Meeting 1762, 11/5/87
Subject:
FWP-2-SOP-8.1A Revision 11, TCN llB FFP-2-SOP-8.1 Revision 11, TCN llB-E
Description:
Rerouted the boron injection tank (BIT) bypass valve leakage to the boron injection surge tank. his configuration involved the opening of BIT outlet isolation valves 8801A and 8801B. Leakage past the BIT bypass valve hLd allowed relatively cold water to flow into the safety injection piping. Wis is believed to have caused thermal cycling of the safety injection lines. Safety Evaluation: With this configuration, the leakage through the BIT bypass valve will be directed to the boron injection surge tank. Procedural guidance is given to maintain the surge tank level between 45 and 80%. W e safety of the plant is not compromised because the BIT outlet valves are open during an emergency and the BIT surge tank is isolated by a safety injection signal. h e safety injection flowpath would not be affected if draining of the BIT surge tank is in progress when a safety injection signal ic received.
he safety evaluation for 'ICN llB-E considered the effects of leakage past the check valves downstream of the BIT. Wese valves block flow ' from the reactor coolant system into the safety injection lines. Excessive leakage past the check valves would be noticed and corrective action could be initiated. If catastrophic failure of the check valves occurred, a loss of coolant accident would exist. A safety injection signal would be initiated automatically. We loss of coolant accident would be terminated since the piping upstream of the check valve would be subjected to a pressure greater than the reactor coolant system preuure. PORC l'eview: PORC Meeting 1794, 12/29/87 PORC Meeting 1796, 12/31/87
Subject:
INP-2-SOP-16.1 Revision 8, TCN 8A
Description:
Defeated the low service water dilution flow trip of 2-BD-RCV-023B during Mode 5 or Mode 6 operation. Normally, this valve will trip closed on low service water dilution flow rate thus isolating the steam generator blowdown (SGBD) path to the environment. Safety Evaluation: Service Water dilution flow is continuously recorded for each unit in the Main Control Room. SGBD releases will be calculated and authorized on a batch release permit to ensure that the total dilution flow from both units is sufficient to ensure that Technical Specification dose limits are not exceeded. SGBD will be caution tagged to require that blowdown be secured if the minimum dilution of the release permit is not met. %e l total dilution flow will be logged every 15 minutes. Defeating the low service water flow trip does not affect the ability of 2-BD-RCV-023B l to close if a high activity level is detected in the SGBD flowpath. PORC Review: PORC Heeting 1750, 10/17/87
Subject:
INP-2-SOP-27.0 Revision 16, KN 16B
Description:
Isolated the normal hydrogen supply to the Auxiliary Building and provided a supply from the main generator hydrogen manifold. Safety Evaluation: he normal volume control tank (VCT) purge will be stopped to limit the amount of make up hydrogen required. With VCT purge stopped, the alternate hydrogen supply will provide sufficient pressure in the VCr for reactor coolant pump seal operation as well os hydrogen make up to replace leakage and hydrogen lost to oxygen scavenging. PORC Review: PORC Meeting 1721, 8/27/87 subject: FNP-2-SOP-31.0 Revision 12, K N 12D
Description:
Provided a backup instrument air supply via hoses for systems that have air-operated components requiring continuous operation during refueling outages. W e normal instrument air supply to containment is required to be isolated for local leak rate testing and some maintenance evolutions on the containment boundary air header isolation valves, his function is normally performed by attaching nitrogen bottles to selected components. Safety Evaluation: Normal system design of instrument air provides a two-inch line with automatic isolation valves which close on a Phase B signal or high penetration room pressure. he backup instru.aent I air supply will isolate on a Phase B signal but it will not isolate on a high penetration room pressure. Since the unit will be in either Mode 5 or 6, the likelihood of receiving a high l penetration room signal is remote. Any leakage at one of the hose connections or via one of the hoses of such magnitude to affect the compressed air system can be isolated readily from outside the containment. PORC Review: PORC Meeting 1761, 11/3/87 l
A 8 .
Subject:
FNP-0-SOP-34.0. Revision 3, TCN 9A
Description:
Isolated the normal hydrogen supply to the Auxiliary Building and installed a tanporary supply from a compressed hydrogen bottle. We hydrogen is supplied through a special adaptor installed in place of the bonnet of valve N1G22v029A. his can be done in order to search for leaks in the hydrogen supply piping. Safety Evaluation: he normal volume control tank (VCT) purge will be stopped to limit the amount of makeup hydrogen required. %e temporary hydrogen supply will provide sufficient pressure in the VCT for reactor coolant pump seal operation as well as hydrogen makeup to replace leakage and hydrogen lost to oxygen scavenging. Since the normal supply has a much greater cape:ity, a leak in the temporary supply will not create a hazard more severe than a leak in the normal supply. PORC Review: PORC Meeting 1721, 8/27/87
Subject:
PNP-2-SOP-37.1 Revision 10, 'ICN 10C
Description:
Provided an alternate means of cooling the battery charger rooms. h is was needed after it was recognized that the battery charger room coolers were supplied service water fran the incorrect train. his TCN addresses the possibility of l losing one train of service water. Some ventilation duct work may have to be removed to I enhance cooling to the battery charger rooms. Safety Evaluation: Removing some ventilation duct work would enhance cooling to the battery charger rooms. his would be done only in the event of the loss of one train of service water. PORC Review: PORC Meeting 1725, 9/3/87 i I l l
Subject:
INP-2-SOP-50.0 Revision 15, TCN 15A
Description:
Deleted the requirement of running a reactor coolant drain tank (RCUT) pump continuously. Safety Evaluation: his change will reduce wear on the RCDT pumps and provide for greater reliability. System parameters are provided at the control station for the RCDT system and are normally monitored and logged approximately every four hours. Also, a high pressure alarm and a high tank level alarm are provided. When required, one pump can be started to maintain parameters in the required , ranges. t PORC Review: PORC Meeting 1668, 4/30/87
Subject:
FNP-2-SOP-50.4 Revision 9
Description:
Deleted the requirement to backflush or fluff a demineralizer after new resin has been added. Safety Evaluation: he existing requirement uses water in the Spent Resin Storage Tank to fluff the new resin. Deleting this requirement will increase both the effectiveness and the life of the resin by not exposing it to contamination before it is used to process fluid. PORC Review: PORC Meeting 1698, 7/9/87 l
Subject:
MD 86-1649
Description:
Changed the setpoint for the high letdown flow alarm. The previous alarm setpoint was 120 gpm. h e revised setpoint is 135 gpm. Safety Evaluation: he letdown flow rate is currently 128 gpm with two letdown orifices in service. mis is greater than the previous alarm setpoint. Changing this alarm setpoint will not affect operation of the unit under normal or accident conditions. PORC Review: PORC Meeting 1620, 1/7/87
Subject:
MD 87-1806
Description:
Installed jumpers to allow the operation of the 2A radioactive waste area exhaust fan with the supply fan out of service. W e jumpers were removed when the supply fan was returned to service. Safety Evaluation: During this temporary minor departure from plant design, the boric acid tank area temperature was monitored to ensure the temperature remained greater than 65'F while the exhaust fan was being run. Also, a caution tag was placed on the exhaust fan control switches requiring them to be shutdown when the outside air temperature was less the 20*r. Appropriate doors between the radiation control area (RCA) and the outside air were opened to allow ventilation. l PORC Review: PORC Meeting 1793, 12/23/87 I l l l
Subject:
Unit 2 Cycle 6 Reload Safety Evaluation (RSE)
Description:
Provided the design for the Cycle 6 core. Revision 0 of the RSE was reviewed on 10/1/87 and several questions were raised. Revision 1 answered these questions and documented the basic cycle 6 design. As described in Revision 1 of the RSE, the Cycle 6 reload design is based on the cycle 5 end-cf-life burnup within a range of 15,600 to 17,300 MND/MIU. A total of 25 Region-6, 68 Region-7, and 64 fresh Region-8 fuel assemblies are used in the design. A total of 656 fresh Wet Annular Burnable Absorbers (WABAs) ace used in clusters of 4, 12, and 16. We Region-8 assemblies differ from the previous design in that they have Reconstitutable Top Nozzles (RINs), a modified fuel rod end plug, and a 4g fuel pellet hold-down spring. The burnable absorber rods differ from past cycles in that they incorporate annular aluminum oxide-boron carbide (A1203 - B4C) absorber pellets contained within two concentric Zircaloy tubes with water flowing through the center tube instead of the current stagnant air-filled centre.1 tube and outer stainless steel clad containing borosilicate glass with B203 absorber. Subsequent to the issuance of the Safety Evaluation, Cycle 6 was redesigned due to the undesirability of reinserting assemblies with damaged grid straps, he damaged grid straps were observed during the EDC5 fuel inspection. We , redesign replaced damaged assembly S-06 with assembly S-55 and damaged assembly S-33 with assembly S-05. Since assemblies S-55 and S-05 were both reinserts, their core positions were replaced by assemblies S-52 and S-46, respectively. Assemblies S-52 and S-46 were l originally planned to be discharged to the spent i fuel pool, he revised Cycle 6 reload design was given in Revision 2 of the RSE. his safety evaluation was applicable to operational Males 5 and 6 only. l l
Revision 3 of the RSE provided the safety evaluation of the redesigned Cycle 6 reload and the evaluation for all operational modes. Safety Evaluation: Based on the RSE and the analyses performed by Westinghouse and Southern Company Services, Inc., this reload does not involve any unreviewed safety questions. PORC Review: PORC Meeting 1741, 10/8/87 (Rev. 1) PORC Meeting 1752, 10/22/87 (Rev. 2) PORC Meeting 1767, 11/12/87 (Rev. 3)
Subject:
FNP Emergency Plan
Description:
'ICN 12A changed the requirement for the anergency Director to possess a Senior Reactor Operator's license. 'ICN 128 alleviated the responsibility of the Public Information Site Coordinator (PISC) to activate the Corporate Conununications Department Emergency Organization. Instead, this responsibility was delegated to the Public Information Emergency Coordinator (PIEC) and the Public Information Activation Assistant. 'IEN 12C replaced the NOAA Weather Alert Radio System for Alert and Notification with a new tone alert system.
Safety Evaluation: With regard to TCN 12A, persons who may serve as Emergency Director must meet the following requirements: (1) Possess the training and experience required for their position as delineated in ANSI 18.1-1971; (2) occupy a position which ht been designated by the General Manager - Nuclear Plant; and (3) have completed Emergency Director Training. 'Ihe Emergency Director Training meets the requirements of NUREG-0654 and 10 Cat 50 Appendix E. With regard to
'ICN 128, the existing plan may cause the PISC to be unduly delayed in his attempt to travel to the site; therefore, the changes will ensure that the PISC will be available to travel to the site in a timely manner should the Emergency Operations Facility be activated. With regard to 'ICN 12C,
the new system will add confidence that the public inside the ten mile Emergency Planning Zone (EPZ) and outside any siren-covered zones will keep their radio in service, thus ensuring a more reliable alert and notification system. (Note: A significant percentage of the public who were issued NOAA weather alert radios were taking their radios out of service due to excessive activaticns from NOAA to notify of severe weather conditions). PORC Review: PORC Meeting 1670, 5/5/87 (KN 12A) PORC Meeting 1723, 9/1/87 (K N 128) PORC Meeting 1724, 9/2/87 ('Im 12C)
Subject:
Safety Evaluation: MOV Backseating Condition
Description:
Revised FSAR Table 6.2-32 to show a new valve closure time for the auxiliary feedwater stop-check valve. This FSAR table previously showed a valve closure time of ten seconds for i these valves. We new closure time is 14 seconds. Safety Evaluation: h e breakers for the motor operators of these valves are required to be locked open while the unit is operating in Modes 1,2 or 3. The valves do not operate automatically in response to any abnormal or accident conditions. Therefore, an increase in the valve stroke time will not affect the safety analysis. I PORC Review: PORC Meeting 1795, 12/31/87 I l l l l
NCrrE: h e following changes were incorporated into the FSAR as part of Revision 5, July 1987.
Subject:
Safety Evaluation: Upr'. ate of FSAR Section 9.4.7.4
Description:
Deleted the requirement for testing those components in the Diesel Building heating and ventilation system which are normally in use. Safety Evaluation: his change does not decrease the reliability of the heating and ventilation system at the Diesel Building. PORC Review: PORC Meeting 1694, 6/30/87
Subject:
Safety Evaluation: FSAR PC 86-15
Description:
Revised some of the specifications for the water treatment system which are specified in FSAR section 9.2.8.2. %ese specifications had been stated incorrectly. Safety Evaluation: This is a change to the text of the FSAR to correct errors in the description of a nensafety related system. No modifications were made to any equipnent. PORC Review: PORC Meeting 1694, 6/30/87
Subject:
Safety Evaluation: FSAR Section 8.3.1.1.4
Description:
Corrected section 8.3.1.1.4 of the FSAR. he text of this section had disagreed with the information in FSAR figures 8.3-23 and 8.3-24 and the "as-built" condition of the equipment involved. he corrected information deals with the backup power sources of the 120 volt AC vital instrumentation power system. Safety Evaluation: he FSAR figures are correct and reflect the "as-designed" and "as built" corditioa of the equipment. Correcting the text is administrative in nature. PORC Review: PORC Meeting 1694, 6/30/87
Subject:
Safety Evaluation: FSAR PC 86-08
Description:
Corrected a discrepancy concerning the location of reactor coolant pump (RCP) vibration displays. RCP vibration displays are located in the electrical penetration room and there is an associated annunciator in the main control room, FSAR section 5.5.1.2(c) had indicated that RCP vibration is displayed in the control room. Safety Evaluation: his change provided a needed correction to the FSAR. PORC Review: PORC Meeting 1694, 6/30/87
Subject:
Safety Evaluation: FSAR PC 86-10
Description:
Corrected a discrepancy in SAR c section 6.5.5. his section had stated that the auxiliary feedwater flow indicators were environmentally qualified. We flow transmitters associated with these indicators are environmentally qualified but the flow indicators themselves are not environmentally qualified. Safety Evaluation: his change provided a needed correction to the FSAR. he indicators are in a mild environment and are not required to be environmentally qualified. PORC Review: PORC Meeting 1694, 6/30/87
Subject:
Safety Evaluation: FSAR Sections 8.3.1.1.2 and 8.3.1.1.6
Description:
Updated FSAR sections 8.3.1.1.2 and 8.3.1.1.6 to take into account additions to the plant's electrical system. Safety Evaluation: his provided a needed correction to the FEAR. PORC Review: PORC Meeting 1694, 6/30/87 . l I l
Subject:
Safety Evaluation: Containment Free Volume
Description:
Revised FSAR paragraph 3.8.1.1.a to remove a discrepancy between this paragraph, FSAR Table 6.2-1, and Technical Specification paragraph 5.2.1.g. FSAR paragraph 3.8.1.1.a had used a larger value for free volume of the containment than the other two sources. Safety Evaluation: Using the lower value of containment free volume is conservative and consistent with the other sources. PORC Review: PORC Meeting 1694, 6/30/87
Subject:
Safety Evaluation: New Appendix 9B to FSAR Which Incorporated the FPPR
Description:
Incorporated the fire protection program into the FSAR as requested by Generic Letter 86-10. We fire protection program had been in a separate document called the Fire Protection Program Reevaluation. We program iF now in FSAR Appendix
- 98. Appropriate additions zac deletions were made to ensure that the new appendix is 11: the proper format and meets applicable requirements. In addition, cable insulation combustible loadings were revised as required.
Safety Evaluation: Wese changes do not decrease the effectiveness of the fire protection program. Changes to the combustible loading were evaluated and found to be acceptable. PORC Review: PORC Meet...g 1691, 6/25/87 1 l l i n
Subject:
GO-EIP-101, Rev. 4, Rev. 5
Description:
his proc @re delineates the organization to be implemented by the general office in support of an emergency. Safety Evaluation: Revision 4 to the procedure reflects management reorganization within Nuclear Generation. Revision 5 to the procedure indientes title and facility name changes.
Subject:
GO-EIP-102, Rev. 3, Rev. A
Description:
his procedure delineates the j organization to be implemented and facilities to be utilized by the corporate communication department in support of an emergency. Safety Evaluation: Revision 3 to the procedure adds a OA records section and reflects several Corporate Comunication title changes. Revision 4 to the procedure centralizes the activities of the Corporate Comunications Department into one procedure. _ . _ ~ _ _ _ _ _ _ - - _ _ _ _ . _- ._- - __ . ..__ - _ - .. _ _ - _ -,. -
Subject:
GO-EIP-111, TCN-11A, 'IEN-11B, Rev.12, Rev. 14, TCN-14A, Rev. 15, Rev. 16
Description:
'Ihis precedure delineates the steps necessary to activate the Nuclear Generation Department Emergency Organization and provide notification of the ovent.
Safety Evaluation: Change TCN-llA adds Environmental Affairs to the call list in the event of a transformer fire which releases PCBs. Change 'ICN-llB adds a table for emergency coordinator notification in the event of a waste transportation accident not requiring EOF action. Revision 12 to the procedure incorporates organizational changes in the offsite Technical Support Organization which combines the duties of the Manager of Nuclear Engineering and Technical Support and the Manager of Nuclear Operations and Administration into the position of General Manager - Nuclear Support. These combined offsite support responsibilities are divided among four managers reporting to the General Manager - Nuclear Support. Phone nu
- r changes were also made.
Revision 34 to the procedure updates the personnel and phone number listings. Change TCN-14A to the procedure involves I phone number changes only. Revision 15 to the procedure involves phone number changeu. title changes, personnel changes, procedural clarification changes, changes to reflect new EOC equipment, and changes to reference the use of the new helicopter. Additionally, a change is incorporated in the EOF Activation Checklist which will prompt the Emergency Coordinator to call the plant and request a list of available plant personnel to be used to supplement EOF staffing. Revision 16 updates phone number, position titles, and personnel changes. Procedural clarifications were made and a step for the EC to notify Alabana Radiological Health was deleted because FNP-EIP-26 calls for the Emergency Director to take this action.
Subject:
GO-EIP-112, Rev. 6, Rev. 6A, Rev. 7
Description:
his procedure provides instructions for establishing conununications between emergency support personnel in transit and personnel located at company facilities. Safety Evaluation: Revision 6 to the procedure incorporates changes made to the Nuclear Generation management organization. Revision 6A to the procedure involves phone and radio number changes. Revision 7 to the procedure involves phone number changes, radio call number changes, and title changes.
Subject:
GO-EIP-114, Rev. 6
Description:
he purpose of these procedure is to provide guidance in the coordination and distribution of news releases in an emergency condition. Safety Evaluation: Revision 6 to the procedure involves changes to agency and personnel titles, adding the Alabama Bureau of Radiological Health to the list for notification of news release content prior to release and adding the Public Information l Emergency Coordinator as an approval for news releases.
Subject:
GO-EIP-116, Rev. 3, Rev. 4 Descriptica: his procedure provides guidance for initially staffing the EOF or changing shifts in the EOF during an emergency. Safety Evaluation: Revision 3 to the procedure incorporates changes in the Nuclear Generation management organization, l Revision 4 to the procedure involves title changes and procedure reference changes to ( ' make the Dose Assessment Director turnover more effective. l l l
Subject:
GO-EIP-ll7, Rev. 5
Description:
his procedure delineates the Emergency Operations racility administrative activities. Safety Evaluation: Revision 5 to the procedure involves phone number and title changes.-
Subject:
GO-EIP-ll8, Rev. 4, Rev. 5
Description:
his procedure delineates the steps necessary to activate the Corporate Communication Department Emergency Organization and provide notification of an event. s Safety Evaluation: Revision 4 to the procedure involves title changes, phone number changes, and procedural format changes. Revision 5 to the procedure incorporates an Activation Assistent, revises methodology to include the Activation Assistant, and adds necessary checklists for emergency positions.
Subject:
GO-EIP-119, Rev. 3
Description:
his procedure delineates the criteria and authority for relocating the EOF from the ' onsite E0r to the Wiregrass District Office. Safety Evaluation: Revision 3 to the procedure includes changes to the equipment list and details of room layouts.
Subject:
GO-EIP-121, Rev. O
Description:
his procedure provides guidance on the activities and responsibilities for Corporate Connunication emergency positions. Safety Evaluation: his procedure is a new procedure that will be used to provide guidance on the necessary activities and responsibilities for several Corporate conmunication emergency positions.
,,_r - . - -, - - - ..- ,
)
Subject:
GO-EIP-131, Rev. 7, Rev. 8 DescLiption: his procedure establishes the actions to be taken to ensure the operational readiness of EOC equipment and supplies. Safety Evaluation: Revision 7 to the procedure incorporates changes within the Nuclear Generation management organization. Revision 8 to the procedure involves the changing of a file number.
Subject:
GO-EIP-132, Rev. 5
Description:
his procedure establishes guidance for the conduct of drills and exercises to maintain emergency preparedness. Safety Evbluation: Revision 5 to the procedure incorporates changes within the Nuclear Generation management organization.
Subject:
GO-EIP-134, Rev. 6 Description his procedure delineates the responsibilities for conducting emergency plan training for corporate personnel. Safety Evaluation: Revision 6 to the procedure incorporates changes within the Nuclear Generation management organization.
Subject:
GO-EIP-135, Rev. 3, Rev. 4
Description:
his procedure delineates responsibilities for Emergency Plan review and revision. Safety Evaluation: Revision 3 to the procedure incorporates changes within the Nuclear Generation management organization. Revision 4 to the procedure involves title changes only.
Subject:
Co-EIP-136, Rev. 3
Description:
his procedure defines the responsibilities of the Corporate Comnunication Department regarding weather alert radio distribution and maintenance. Safety Evtluation: Revision 3 to the procedure incorporates changes within the Nuclear Generation management organization.
Subject:
GO-NG-2, Rev. 3, Rev. 4
Description:
his procedure describes the organization and responsibilities of the General Office Nuclear Support section. Safety Evaluation: Revision 3 to the procedure is an organizational change reflecting combined duties of:
- Manager Nuclear Engineering and Technical Support; and
- Manager Nuclear Operation and Administration into General Manager Nuclear Support.
Revision 4 changes the revision number only. he existing revision number 3 was inadvertently used instead of revision 4.
Subject:
GO-NG-5, Rev. 3
Description:
his procedure describes organizational functions and responsibilities for conducting an environmental monitoring program. , Safety Evaluation: Revision 3 to the procedure is an editorial change j to reflect the change in title from Plant Manager to General Manager - Nuclear Plant. Subjo:;' : GO-NG-44, Rev. 3 l
Description:
This procedure estab7.ishes the guidelines to be used by Nuclear Support for the coordination of inservice inspectiori activities in support of Farley Nuclear Plant. Safety Evaluation: Revision 3 to the proceitre implements changes to show specific due dates, the appropriate NRC addressee, and the re-{uirement source for each submittal. h e duties of the Senior Project l Engineer - ISI were also clarified to include the ! review and amendment of the: FSAR and Technical l Specifications requirements as required. i l
i' ~. - .. Alabama Power Company 600 North 18th Street Post Office Bo. 2641 Dirmingham. Alabama 352910400 Telephone 205 250-1835 R. P Mcdonald m Senior Vice President Alabama Power the southern eJectrc system 10CFR50.59 March 31, 1988 Docket No. 50-364 U. S. Nuclear Regulatory Conunission ATIN: Doctanent Control Desk Washington, DC 20555 Gentlemen Attached for your review is the annual report required by 10CFR50.59 for 1987. 'Ihis report summarizes changes to the plant performed in accordance with the provisions of 10CFR50.59 for Joseph M. Farley Nuclear Plant Unit 2. If you have any questions, please advise. 1 Respectfully ubmitted, f , I L/ _, l RPW REM: dst-D1.46 Attachments cc: Mr. L. B. Long l Dr. J. N. Grace l Mr. E. A. Reeves Mr. W. H. Bradford 1
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