ML17059A088
| ML17059A088 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 12/31/1993 |
| From: | NIAGARA MOHAWK POWER CORP. |
| To: | |
| Shared Package | |
| ML17059A087 | List: |
| References | |
| NMP2L-1448, NUDOCS 9311040065 | |
| Download: ML17059A088 (96) | |
Text
Enclosure to NhiP2L 1448 NXNE MILE POPtT - UVlT 2 SAFETY EVALUATIONS YREPORT 1993 Docket No. 50-410 License No. NPF-69 93110400hS 931029 PDR ADOCK 05000410 K" PDR
Safety Evaluation Summary Report Page 1 of 93 Safety Evaluation No.: 89-015, Rev. 0 & 1 Implementation Document No.: Mod. PN2Y88MX194 USAR Affected Pages: Sections 1.2, 8.3, 9.2, 9.4, 9.5, 9A.3, 12.3 Systems Various Title of Change: Cafeteria Building Description of change:
This modification constructed a two-story, steel-framed (approximately 7,500 sq. ft.) cafeteria building with its own foundation and structural elements adjacent to the west wall of the turbine building between the existing -chill water building and the foam room. Power feed for the new cafeteria panel 2NJS-PNL302 was from load center 2NJS-US3, located in the turbine building at elevation 277'. To enable construction of the cafeteria and its foundation, modifications to existing plant structures and components were necessary in the yard area, foam room, I&C shop, passageway, turbine building, and the chill water building. Affected systems included fire protection water, domestic water, HVAC, and lighting. The 6-inch block wall on the east end of the cafeteria building (side facing the Unit 2 turbine building) was filled with concrete, thereby providing radiation protection from gamma radiation resulting from hydrogen injection.
Safety Evaluation Summary:
This modification resulted .in the addition of a nonsafety-related structure. Evaluations were performed to determine potential effects on safety-related systems and structures, ALARA concerns, fire protection, the site flood analysis, and electrical power supply. The evaluations concluded that there were no adverse effects resulting from this modification.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 2 of 93 Safety Evaluation No.: 90-039 Implementation Document No.: Mod. PN2Y89MX083 USAR Affected Pages! Section 3.7A Systems Seismic Monitoring Title of Change! Permanent Relocation of Accelerographs 2ERS-PAC2C and PAC2B Description of Changes This modification relocated seismic monitor 2ERS-PAC2C from the recirculation pump motor to the reactor pedestal. This modification also relocated seismic monitor 2ERS-PAC2B to a location on the same CSH line but closer to the CSH penetration at the primary containment wall. Xt had been determined that both-units failed, in part, due to high background vibration.
Both of the new locations were selected based on anticipated low background vibration.
Safety Evaluation Summary:
This modification established a new location for a triaxial peak accelerograph to maintain compliance to Regulatory Guide 1.12 and enabled the instrument to accurately monitor seismic activity.
The relocation of this instrument will not affect the operation of any safety systems or the safe operation or shutdown of the plant. The placement of the accelerograph on the reactor pedestal does not conflict with the original design intent. The new instrument requires calibration in accordance with the manufacturer's recommendations. The new location on the reactor pedestal allows the instrument to measure seismic movements that can be compared to the analytical seismic response without the seismic traces being masked by nonseismic traces from background vibration. Amendment No. 39 to the facility operating license revised Technical Specification Tables 3.3.7.2-1 and 4.3.7.2-1 to reflect the relocation of seismic monitor 2ERS-PAC2C.
2ERS-PAC2B The relocation of accelerograph 2ERS-PAC2B to a new location on the same primary containment CSH line as the original location did not change the USAR or Technical Specifications. The new location is expected to minimize the accelerograph's exposure to both background vibration and personnel contact.
Safety Evaluation Summary Report Page 3 of 93 Safety Evaluation No.: 90-039 (cont'd.)
Safety Evaluation Summary! (cont'd.)
Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.
this
Safety Evaluation Summary Report Page 4 of 93 Safety Evaluation No.: 90-076 Implementation Document No.: Dwg. No. 001.6550-076-092,093; EDC 2E00528 USAR Affected Pages: Section 9.1 System: Spent Fuel Pool Cooling and Cleanup (SFPC)
Title of Change: SFPC Filter/Demin. Effluent Flow Switch Setpoint Change Description of Change:
Flow switch 2SFC-FAL47A/B initiates a system trouble alarm and starts filter/demineralizer holding pump 2SFC-P4A/B automatically when the flow in SFPC filter/demineralizer FLT1A/B drops below the low flow setpoint. This change increased the low flow setpoint from 360 gpm to 700 gpm to allow for proper instrument calibration.
Safety Evaluation Summary:
Raising the low flow setpoint from 360 gpm to 700 gpm does not adversely impact nuclear safety. Flow switch 2SFC-FSL47A/B continues to perform the alarm and start functions of the holding pump in the same manner as the existing design. The normal flow rate seen by 2SFC-FT47A/B is either 1200 gpm or 2400 gpm. The difference between the normal flow rate and the low flow setpoint is adequate to avoid spurious actuation.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 5 of 93 Safety Evaluation No.: 90-081 Implementation Document No.: EDC 2M00306, FDDR KG1-0866-1 USAR Affected Pages: Sections 3.9B, 4.6 Systems "
Control Rod Drive (CRD)
Title of Change: Addition of Control Rod Drive (CRD) Model 7RDB144EG001 Description of Change:
New CRDs (Model No. 7RDB144EG001) were purchased as replacement parts to be installed in the reactor 'vessel as part of the routine CRD Maintenance Program. The design of the new CRD incorporates the following design changes:
The material for the strainer, inner filter, and outer filter was changed from stainless steel type 304 to type 304L to enhance their resistance to Intergranular Stress Corrosion Cracking (IGSCC).
2 ~ The design of the uncoupling rod was changed from two pieces (uncoupling rod and tube) welded together, to a single piece made of solid bar. A segment of the uncoupling rod has a triangular cross-section to assure proper installation in the center hole of the spud, thereby eliminating the possibility for human error during the assembly of the CRD.
3 ~ The set screw plug which contained a single through-hole coolant passage was replaced with a cooling orifice containing eight side holes to reduce the possibility of coolant flow blockage by foreign material. This design change results in a small increase of approximately 0.02 gpm in the coolant flow through each CRD (3.7 gpm for the total of 185 CRDs) .
- 4. The position indicator probe is not included as a part of the new CRD, but rather is supplied separately.
In addition, a smaller number of the new drives will be subjected to the 5-year maintenance life tests due to the lower production volume expected for this model.
The CRDs are used for positioning the control rods in the reactor core. Changes described in this safety evaluation will not have any adverse impact on either the normal or the safety-related CRD functions, including scram capability or scram speed of the CRD.
Safety Evaluation Summary Report Page 6 of 93 Safety Evaluation No.: 90-081 (cont!d.)
Safety Evaluation Summary:
CRD Model 7RDB144EG001 has been reviewed against "BWR Scram Discharge System Safety Evaluation," issued by the NRC in December 1980, and found to have no impact on the evaluation or conclusions.
The slight increase in flow caused by the design change of the cooling orifice is within hydraulic capacity of the control rod drive hydraulic system (RDS). Performance of operating procedures for the RDS will verify charging requirements are met.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 7 of 93 Safety Evaluation No.: 90-099 Implementation Document No.: Simple Design Change SC2-0006-90 USAR Affected Pages: Sections 1.2, 4.6, 9.3, 12.3 System! Control Rod Drive (RDS)
Title of Change: Control Rod Drive Maintenance Room Description of Change:
The following changes were made in the CRD maintenance room located in the secondary containment.
The existing CRD flush tank was removed and replaced with a new, shielded flush tank that includes an effluent filtration system, pump, and bottle accumulator. The tank was also relocated within the room.
- 2. The existing flush tank drain line and valves were removed and the piping was capped.
- 3. The control cabinet for the ultrasonic cleaner was permanently mounted on the north wall of the room.
Safety Evaluation Summary:
These changes to the CRD maintenance room will greatly improve CRD handling, maintenance and storage for refueling outages, and reduce the possibility'f 'creating high radiation areas in the CRD maintenance room and on elevation 261'f the secondary containment. The CRD maintenance room equipment is not required to support safe shutdown of the plant. The changes addressed in this safety evaluation are nonsafety related and do not adversely affect the ability of any systems or components important to safety to perform their safety function.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 8 of 93 Safety Evaluation No.: 90-103 Implementation Document No.: Mod. PN2Y86MX085 USAR Affected Pages: Section 9.3 System e~ Instrument Nitrogen (GSN)
Title of Changes Control Room Nuisance Alarms CEC601505, CEC601506 Description of Change:
The nitrogen system trouble alarm (annunciator window 601506) and the primary containment purge temperature low alarm (annunciator window 601505) activate under low process temperatures to indicate a problem with the trim heaters and/or electric vaporizers. Nuisance alarms were being generated when there was no or low nitrogen flow because the source of the signal for annunciation is located outside (subject to low ambient air temperature), where heat was being lost from noninsulated process piping. To eliminate these nuisance alarms, electric heat tracing/insulation was added on the nitrogen inerting system piping from the trim heaters/electric vaporizers, in the yard (nitrogen tank) area, to a point prior to adjoining the containment purge system, in the standby gas building.
Safety Evaluation Summary!
This change eliminates nuisance alarms by maintaining nitrogen line temperature and reducing heat transfer losses from the process. The alarm logic as designed still provides the functional capability of alarming if the heaters are inoperable.
The ability to maintain the nitrogen temperature above the minimum desirable temperature is also enhanced by this change.
Based on the evaluation performed', it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 9 of 93 Safety Evaluation No.: 90-112, Rev. 3 & 4 Implementation Document No.: N/A USAR Affected Pages: N/A System: Various Title of Change: Office and Storage Facilities in Support of Outages at the Nine Mile Point. Site Description of Change:
As reported in letter NMP2L 1324, dated October 30, 1991, under Safety"Evaluation 90-112, Rev. 1 and 2, temporary office and storage facilities were installed for use by the contractor staff during Unit 2 refueling outages.
Revision 3 to Safety Evaluation 90-112 included an additional temporary facility, a diesel tank storage berm. This storage berm provides environmental protection against spillage of diesel fuel from three fuel storage tankers located within the boundaries of the berm.
Revision 4 to Safety Evaluation 90-112 included a newly-proposed design for a permanent access control building for replacement of the existing temporary radiation protection portal facility.
Safety Evaluation Summary:
These construction activities and temporary, site changes do not adversely impact the site flooding analysis (probable maximum precipitation). Electrical power for the south auxiliary bay will exit portal is taken from the construction power loop and not adversely affect permanent plant power systems.
Based on the evaluation performed, does not involve an it unreviewed is concluded that this safety question.
change
Safety Evaluation Summary Report Page 10 of 93 Safety Evaluation No.: 91-029, Rev. 1 & 2 Implementation Document No.: Simple Design Change SC2-0205-91 USAR Affected Pages! Sections 9.3, 11.3 S'stem e f Of gas (OFG)
Title of Change: Offgas Freeze-Out Dryer Bypass and Drain Line Reroute
. Description of Change:
This design change was twofold. First, to prevent the possibility of organic material from entering the condenser, and thus the reactor, through the condensate system (in the event. of freeze-out dryer refrigerant tube leaks), the freeze-out dryer drain was rerouted to radwaste drain tank 2DFT-TK1B. This involved the addition of a new drain line, and isolation of the existing drain line by closing valve 2OFG-V43 at condenser nozzle 135A, and by closing a new second isolation valve, 20FG-V300.
Secondly, for each freeze-out dryer (20FG-DRYlA, 1B and 1C) inlet process line, a bypass drain line was added to continuously remove entrapped moisture.
Safety Evaluation Summary:
This change will not impact the operation of any safety-related systems or affect the safe operation or shutdown of the plant, and will not impact any equipment associated with the offgas process. Hydrogen detonation design criteria are not impacted.
Offgas drain fluid will be analyzed, and any freon and/or oil, detected, will be processed by an environmentally and if radiologically approved method. New pipe installed meets the original design, specifications and material requirements.
Based on the evaluation performed, changes do not involve an unreviewed it is concluded that these safety question.
Safety Evaluation Summary Report Page 11 of 93 Safety Evaluation No.: 91-034 Implementation Document No.: 'Mod. PN2Y86MX085 USAR Affected Pages: Sections 7.3, 9.3 System: Reactor Water Cleanup, Reactor Plant Sampling Title of Change: Control Rod Nuisance Alarm Windows 602315, 602317, 602318 Description of Change:
Whenever the reactor water cleanup (RWCU) filter/demineralizers are out of service, conductivity goes high due to process fluid stagnation. This causes nuisance alarms since control 'room operators are only concerned with high/low conductivity levels when the associated demineralizer train is in operation.
To eliminate these nuisance alarms, the RWCU filter demineralizer effluent conductivity high/low alarm input was interlocked with the associated filter demineralizer operation logic such that the alarm is inhibited when the demineralizer is not in service.
In addition, the redundant high conductivity alarm input,to the associated filter demineralizer "trouble" annunciator window was deleted by reprogramming the programmable controller associated with each filter demineralizer train and connecting its output to the existing "high/low" annunciator window.
Safety Evaluation Summary:
This modification corrects the nuisance alarms without adversely affecting the RWCU system or the reactor sampling system, and will have no impact on the safe operation or shutdown of the plant.
Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.
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Safety Evaluation Summary Report Page 12 of 93 Safety Evaluation No.c 91-038 Implementation Document No.: EDC 2E10345 USAR Affeoted Pages! Section 10.4 System! Low Pressure Heater Drains (HDL)
Title of Changes Level Indicating Controller ID Changeout Description of Change:
This change reidentified the. fourth point heater water level controllers from 2HDL-LIC4A,B, & C to 2HDL-LIK4A,B & C to correctly identify their form and function. The LIC to LIK designation change correctly identifies that the controllers are capable of providing the level setpoint automatically or manually with operator control as originally specified.
Safety Evaluation Summary:
This change does not affect the design or function of the fourth point heater water level controllers. The controllers were reidentified to properly indicate that they possess the auto/manual transfer capability as originally specified and purchased.
Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.
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Safety Evaluation Summary Report Page 13 of 93 Safety Evaluation No.: 91-042, Rev. 2, 3 & 5 Implementation Document No.: Mod. PN2Y89MX078 USAR Affected Pages: Sections 1.2, 5.2, 10.1, 10.4 System: Oxygen Feedwater Injection (OFI)
Title of Change: Installation of the Oxygen Feedwater Injection System Description of Change:
This modification added the oxygen feedwater injection (OFI) system. The purpose of the OFI system is to maintain 20 to 50 ppb of oxygen in the condensate/feedwater systems. The injection of oxygen in the suction side of the condensate pumps will minimize corrosion and corrosion products released from the condensate/feedwater system materials. The OFI system consists of two parts. The first part is the OFI supply which consists of six oxygen cylinders, an oxygen control manifold, excess flow check valve and relief valve. The second part is the injection portion, which is operated in a manual or manual bypass mode and is composed of isolation and bypass valves, flow controller, solenoid and check valves, backpressure regulator, and oxygen and condensate flow indicators.
Safety Evaluation Summary:
The OFI system does not perform any safety-related functions.
Failure of the system to perform its intended function will not have an adverse impact on the condensate and feedwater systems, nor will it have any impact on the performance, availability and reliability of any other safety-related system. The addition of oxygen in the condensate/feedwater systems will reduce the rate of corrosion and corrosion products released from the system's carbon steel and stainless steel components, and will assist the NMP2 design in meeting ALARA requirements. The added oxygen will not affect the chloride concentration, conductivity, pH, or specific activity in the primary coolant.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 14 of 93 Safety Evaluation No.: 91-043, Rev. 1 & 2 Implementation Document No.: Major Order No. 0546 USAR Affected Pages: Sections 1.2, 2.4 System: N/A Title of Change! New York Telephone Switch Building at Nine Mile Point Unit 2 Description of Change:
The original telephone system on site was inadecjuate to meet the needs of site personnel. A new single switch replaced the two switches previously in use at Units 1 and 2. The new system is housed in a new building outside the protected area, west of the east flood control berm at Unit 2.
Safety Evaluation Summary:
The new single switch facilitates the entire site telephone system as well. as meeting the future of data communication. The new building is not within the direct flow path of flood waters, and thus will have no adverse impact on the probable maximum precipitation (PMP) flood study.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report, Page 15 of 93 Safety Evaluation No.: 91-052 Implementation Document No.: Mod. PN2Y86MX085 USAR Affected Pages: Figure 9.2-2 Sh. 5 System: Service Water (SWP)
Title of Change: Control Room Nuisance Alarm Windows 601134, 601222 Description of Change:
When power to one of the SWP strainers (2SWP*STR4A,B,C,D,E,F) is disconnected, a motor overload signal is sent to the control room annunciator windows 601134 and 601222 causing false and unwanted nuisance alarms.
This modification eliminated the nuisance alarms by interlocking the motor overload circuits with the loss-of-power circuits for each of the strainers so that the motor overload alarm is sent to the control room annunciators only when power is on and the motor is actually overloaded.
The effects of Safety Evaluation 91-052 were incorporated in Figure 9.2-2 Sh. 5, USAR Revision 4.
Safety Evaluation Summary:
The design change associated with the SWP strainer motor control circuit will not compromise the design for the existing bypass/inoperable status indication and annunciation for these circuits. The electrical separation between divisional and nondivisional circuits will be maintained, and the design associated with the SWP strainer motor control circuit change such that a single failure will not will be "fail-safe" system's compromise the SWP ability to perform it's protective function. The SWP strainers'peration and function will not change.
This modification corrected'he nuisance alarms without adversely affecting the SWP system and will have no impact on the safe operation or shutdown of the plant.
Based on the evaluation performed, it is concluded that this safety question.
change does not involve an unreviewed
Safety Evaluation Summary Report Page 16 of 93 Safety Evaluation No.: 91-054, Rev. 1 Zmplementatfi.on Dooument No.: Mod. PN2Y86MX085 USAR Affected Pages: Section 9.2 System: Service Water (SWP)
Title of Change!. Control Room Nuisance Alarm Windows 601115, 601218 Description of Change:
Each of the service water pumps, 2SWP*P1A, 1B, 1C, 1D, 1E, 1F, have a low suction pressure alarm to notify the operators of a low suction pressure condition. When a pump is taken out of service at the present time for maintenance, the pressure switch senses a low suction pressure and sends a signal to the control room annunciator windows 601115 or 601218, causing false and unwanted nuisance alarms.
To eliminate these nuisance alarms, the service water pump low suction pressure circuits were interlocked with the associated service water pump control switch circuits to inhibit the alarm input when the switch is in the PULL-TO-LOCK or STOP positions.
The switch acts as a permissive to allow the alarm to perform as required when placed in the START, NORMAL-AFTER-START, or NORMAL-AFTER-STOP positions.
Safety Evaluation Summary:
The operators are concerned with the service water pump low suction pressure only when the associated pump is in operation or in the standby mode. This change eliminates the actuation of this nuisance alarm when a maintenance activity is performed on a service water pump train resulting in this condition. The service water pumps'peration and function do not change.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question. F
Safety Evaluation Summary Report Page 17 of 93 Safety Evaluation No.: 91-055 Implementation Document No.: EDC 2F00342 Procedures N2-0P-48, N2 CSP Q620g N2 CSP D621g N2-CSP-W622 USAR Affected Pages: Section 9.5 System! Auxiliary Boiler Chemical Feed (ABH)
Title of Change: Auxiliary Boiler Chemical Feed System Description of Change:
This change revised the description of the auxiliary boiler chemical feed system to accurately reflect the existing plant operational and configuration status. The previous description stated that the system utilized sodium sulfite in tank 2ABH-TK1 and sodium hydroxide in tank 2ABH-TK2. This change revised the description to reflect that tank 2ABH-TK1 contains sodium phosphate and tank 2ABH-TK2 contains sodium sulfite. The corresponding pumps for these tanks have also been correctly described consistent with the tank contents.
Safety Evaluation Summary:
The auxiliary electric boiler system is nonsafety related and is not required to support safe shutdown of the reactor or to perform in the operation of reactor safety features. This documentation change reflects the actual operational configuration of the auxiliary boiler chemical feed system and is consistent with industry standards and vendor recommendations.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 18 of 93 Safety Evaluation No.: 91-062 Implementation Document No.: Temporary Mod.91-056 USAR Affected Pages: N/A System: Fire Protection Water (FPW)
Title of Change: Temporary Closure of Valve 2FPW-V215 Description of Change:
This temporary change allowed the extended closure of fire protection water supply valve 2FPW-V215 and jumpered out the trouble indication, computer alarm, and nuisance alarm signal initiated from this valve's position switch (2FPW-ZS415). Valve 2FPW-V215 supplies water to the sprinkler piping on main transformer 2MTX-XM1B (fire zone 502SW). It's closure was required for the disassembly of the sprinkler piping to allow removal of the transformer for repair/replacement.
Safety Evaluation Summary:
This temporary change does not affect the safe operation or safe shutdown capabilities of the plant.
Valve 2FPW-V215 supplies water to only one fire zone (502SW); and transformer 2MTX-XM1B is the only component in fire zone 502SW.
The remaining main transformers (2MTX-XM1A/C/D) each have their individual fire zones and water deluge systems controlled by their individual water supply valves. The main transformers are separated from each other by concrete fire walls. The closure of the water supply to fire zone 502SW, therefore, will not place at risk any other components in fire zone 502SW or in any adjoining fire zones.
Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed,safety question.
Safety Evaluation Summary Report Page 19 of 93 Safety Evaluation No.: 91-069 Implementation Document No.: Major Order No. 0545 USAR Affected Pagess Sections 1.2, 2.4 System! N/A Title of Change: Nine Mile Point Unit 2 Site Paving and Drainage Description of Change:
This change regraded and paved the parking lot south of the "P>>
building (an area of approximately 16,000 sq. yds.). Drainage of the swale south of the parking area, running to the east and then to the north, was also improved by lining with geotexal fabric and cobblestone, and the existing culvert under the east service road was abandoned. A 12'-0" 'paved turning lane was also added to Lake Road between the warehouse road and the east service road.
Safety Evaluation Summary:
This change to the parking lots as well as turning lane into the east service road improves the flow of traffic into and out of the plant. Improving the condition of the parking lots eliminates the possibility of personnel injury due to loose rocks and standing water conditions. A review of the flood study calculations determined that this change improves site drainage and has no adverse effect on the probable maximum precipitation (PMP) flood elevation.
Based on the evaluation performed, it is concluded that this safety question.
change does not involve an unreviewed
Safety Evaluation Summary Report Page 20 of 93 Safety Evaluation No.: 91-072, Rev. 1 Implementation Document No.: Mod. PN2Y86MX085 USAR Affected Pages: Section 7.4 System: Reactor Core Isolation Cooling (RCIC)
Title of Change! Control Room Nuisance Alarm Window 601320 Description of Change:
The RCIC turbine exhaust header drain pot is equipped with a level switch (2ICS*LS206) that activates control room annunciator 601320, and provides a signal to open air-operated valve (AOV) 2ICS*AOV110 on high water level simultaneously. With valve 2ICS*AOV110 open, water drains from the drain pot to the reactor building floor drains through AOVs 2ICS*AOV109 and 2ICS*AOV110.
Presently when valve 2ICS*AOV110 opens, the water level quickly drops, the switch resets, the alarm clears, and the valve closes.
This occurs frequently and makes annunciator window 601320 a nuisance alarm that continuously alarms and resets in response to normal events. This modification added two time delay relays to level switch 2ICS*LS206. The first time delay relay keeps valve 2ICS*AOV110 open for a designed period of time. The second time delay relay was connected in the circuit that actuates annunciator window 601320. It allows the actuation of annunciator window 601320 on high drain pot level only after the designated period of time has passed and the level switch has not reset.
Safety Evaluation Summary:
This modification corrects a nuisance alarm without adversely affecting the operation or function of the RCIC system. The revised alarm circuit will alert the operator to the abnormal condition that the water level in the drain pot has not decreased sufficiently once the drain valve has been signalled to open due to drain pot high level. The current logic which closes 2ICS*AOV110 upon opening of steam supply valve 2ICS*MOV120 remains intact and is not being changed. This will assure that 2ICS*AOV110 will be closed upon steam supply through 2ICS*MOV120.
Based on the evaluation performed, it is concluded that this change does not, involve an unreviewed safety question.
Safety Evaluation Summary Report Page 21 of 93 Safety Evaluation No.: '91-074, Rev. 1, 2 & 3 Implementation Document No.: Mod. PN2Y88MX059 USAR Affected Pages: Sections 4.6, 5.4 System: Reactor Water Cleanup (RWCU),
Control Rod Drive (CRD)
Title of Change: CRD to RWCU Pumps Piping Installation Description of Change:
This modification installed a supply of seal injection water from the CRD system to the, seal injection piping of the RWCU pumps.
This results in demineralized quality water being utilized for RWCU pump seal injection, which is a much better quality water than presently used. Recommendation to install this modification, known as "seal purge," was provided to utilities in General Electric (GE) Service Information Letter (SIL) Number 258, Supplement 1. In addition, a 3/4-inch bypass valve was installed for each RWCU pump discharge check valve as recommended in GE SIL Number 258, Supplement 2. The bypass valve can be used for warmup of an idle pump using bleed backflow from an operating pump. The existing seal injection system was not disabled, but was valved out for normal operation. This allows RWCU pump operation with the presently existing seal injection system (after valve realignment) should the CRD system be unavailable for any reason (e.g., during outages).
This change also permanently removed a RWCU system thermal relief valve, 2WCS-RV143, which had previously been temporarily removed under Temporary Modification 92-007 (Safety Evaluation 92-004).
Safety Evaluation Summary:
The new RWCU pump seal injection system is expected to increase seal life.and improve system operability. Radiation exposure of operations and maintenance personnel should be reduced due to longer seal life and improved pump warmup procedures. CRD system design and performance requirements are not impacted and RWCU system design requirements remain fulfilled by this modification.
Piping installed is to ASME and ANSI Code requirements and is seismically supported so as not to impact the operation of any equipment considered important to safety.
Safety Evaluation Summary Report Page 22 of 93 Safety Evaluation No.: 91-074, Rev. 1, 2 & 3 (cont'd.)
Safety Evaluation Summary: (cont'd.)
This modification results in the diversion of up to 16 gpm of CRD flow to the RWCU pump seals. This additional demand will reduce system pressures but pressures will still be sufficient to maintain the minimum scram accumulator pressure without requiring any CRD system adjustments.
The analytical limit for the differential RWCU flow Technical Specification setpoint was increased to allow for the additional unmonitored inlet flow introduced by this modification. Adequate margin exists in the basis for the setpoint such that the analytical value can be increased by 20 gpm without impacting the allowable value or setpoint.
Based on the evaluation performed, involve it unreviewed is concluded that this safety question.
change does not an
Safety Evaluation Summary Report Page 23 of 93 Safety Evaluation No.: 91-077, Rev.', 2, & 3 Implementation Dooument No.: Mod. PN2Y89MX042 USAR Affected Pages: Sections 1.2, 8.3, 9.5, 9A.3, 9B.10 System: Uninterruptible Power Supply (UPS)
Title of Change: Replace 2VBB-UPS1C and 2VBB-UPS1D Description of Change:
The purpose of this change was to eliminate the overloaded conditions on normal UPSs 2VBB-UPSlC and 1D in order to maximize the reliability, maintainability, and performance of these units (reference LER 89-014).
The scope of this modification was to:
- 1. Replace 2VBB-UPS1C and 1D with new equipment to improve reliability and maintainability.
- 2. Shed loads to remove overloaded condition of this equipment by:
- a. Reducing wattage of essential lighting fixtures in turbine and screenwell buildings.
- b. Power those essential lighting fixtures located adjacent to 8-hour battery-pack lights from normal lighting.
- c. Change receptacle power feeds from essential to normal power.
USAR descriptions of the lighting systems have also been revised to more clearly indicate those lighting systems that are available to safely shut down the plant during abnormal or fire events.
The plant emergency and 8-hour battery-pack lighting systems are required for safe shutdown of the plant in the event of an emergency condition evaluated in the USAR. This modification does not alter the present configurations of the emergency or 8-hour battery-pack lighting systems that are required for safe shutdown of the plant.
Safety Evaluation Summary Report Page 24 of 93 Safety Evaluation No.: 91-077, Rev. 1, 2, & 3 (cont'd.)
Safety Evaluation Summary:
The essential lighting system provides partial illumination for certain critical areas of the plant such as control room, relay and computer room, standby diesel generator rooms, emergency switchgear rooms, service water pump rooms, and to and from areas where safety-related equipment is located.
The changes to the plant essential lighting system are still within the design basis of the plant because:
- 1. The minimum required illumination of 0.5 foot-candle for all exit paths is maintained.
2 ~ Wherever the essential lighting is converted to normal lighting, the emergency and/or 8-hour battery-pack lighting is still available for safe shutdown of the plant in the event of a fire in any fire area of the plant or in the event of an accident or transient described in the USAR.
- 3. The safe shutdown capability of the plant either from control room or other areas of the plant (such as remote shutdown room) is not compromised.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page 25 of 93 Safety Evaluation No.: 91-078 Implementation Document No.! Simple Design Change SC2-0141-90, EDC 2F00418 USAR Affected Pages: Section 10.4 Systemo Circulating Water (CWS)
Title of Change: Circulating Water System Conductivity Equipment Description of Change:
Equipment existed at the discharge flume within the cooling tower screenhouse to measure circulating water conductivity. The conductivity element (2CWS-CE103) and associated'quipment (2CWS-CIT103) were no .longer used to calculate blowdown requirements by means of a conductivity measurement. This change removed the conductivity element (2CWS-CE103) and retired in place the transmitter (2CWS-CIT103).
Safety Evaluation Summary:
The circulating water system is not required to effect or support safe shutdown of the reactor or to perform in the operation of reactor safety features. Elimination of the conductivity analysis equipment will not impact the function or the performance of the CWS system.
The design basis water quality of the CWS system is made up of various concentrations of different constituents. Actual operating constituent concentrations in the closed loop are maintained at less than two times the actual lake water concentrations as a result of the constant makeup and a controlled blowdown flow. Circulating water analysis is procedurally controlled by the Chemistry Department on a regular basis. The Chemistry Department has strict procedural controls in relation to blowdown requirements. Copper discharge limits are now the controlling factor in determining blowdown requirements. Conductivity can be determined by the Chemistry Department by means of grab sample when necessary.
Based on the evaluation performed, change does not involve an it unreviewed is concluded that this safety question.
Safety Evaluation Summary Report Page 26 of 93 Safety Evaluation No.: 91-084 Implementation Document No.: Mod. PN2Y89MX094 USAR Affected Pages: Sections 3.9A, 6.2 Systems Containment Atmosphere Monitoring (CMS)
Title of Change: Elimination of 2CMS*SOV25A, B, C, D Description of Change:
The design of the hydrogen/oxygen monitoring system provides for both manual and automatic sampling of primary containment by two fully independent hydrogen/oxygen analyzer trains. This monitoring can be accomplished by manually selecting a sample from five different areas (per division); three from the drywell, and two from the suppression chamber. The sampling point could also originally be automatically selected and sequenced by a cycle timer opening and closing selector valves for the five different areas, alternating between drywell and the suppression chamber.
This modification removed the valve internals from the suppression chamber selector valves, 2CMS*SOV25A, B, C, D, functionally eliminating these valves from the sample path. This change eliminated the capability for automatically selecting and sequencing of sample'ocation, alternating between drywell and suppression chamber. The capability for automatically selecting and sequencing between drywell areas remains.
Safety Evaluation Summary:
The elimination of the selector valves in the suppression chamber does not affect the safety-related function of the hydrogen/oxygen monitoring system. The capability still exists to sample the drywell and suppression chamber, a representative sample is still being taken, and the transport and accuracy of the sample is not affected. Valves 2CMS*SOV25A, B, C, D did not perform a containment isolation function. The isolation function between suppression chamber and drywell sampling that was previously performed by these valves can be provided by the associated CMS system containment isolation valves.
Based on'he evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 27 of 93 Safety Evaluation No.: 91-087, Rev. 1 & 2 Zmplementation Document No.: Mod. PN2Y88MX058 USAR Affected Pages: Sections 3.6A, 5.4, 8.3, 10.1, 10.4 System: Reactor Water Cleanup (WCS)
Title of Change: Feedwater Stratification Reduction of Long-Term Effects Description of Change:
This modification consisted of adding an 8-inch motor-operated globe valve to each WCS line before it joins the feedwater (FWS) line at the thermal tee, and implementation of a new operating procedure to permit direction of all WCS flow through one feedwater line with feedwater flow through two feedwater lines at reactor power levels under 204. This capability will reduce the maximum differential temperature with increased WCS flow, resulting in an increase in the allowable cycles and corresponding life of the pipe. As a minimum, the shifting of the WCS flow to either of the two feedwater lines will double the existing predicted life of the pipe as injection.
it relates to WCS Twenty-two thermocouples were also provided, two on the WCS system and twenty on the FWS system, to monitor the alarm stratification in the feedwater piping thermal tee region.
Safety Evaluation Summary:
'The WCS system is not safety related and is not required for safe shutdown. The new valve operators and all controls are also nonsafety related. All nonsafety-related equipment in
. safety-related areas are seismically supported.
The addition of the valves to lines 2-WCS-008-89-1 and 2-WCS-008-250-1 creates a negligible increase in pressure drop in the system when the valves are full open. When these valves are at other than full open, the WCS system design is bounded by the existing system analysis. Total combined mass flows and enthalpy of feedwater and RWCU water do not change. The changes in WCS flow control occur only at reactor power levels under 204.
The reactor pressure boundary is unaffected by this change since the original containment isolation valves and associated piping remain unchanged. Design temperatures and pressures remain unchanged as do normal operating temperatures.
Safety Evaluation Summary Report Page 28 of 93 Safety Evaluation No.: .91-087, Rev. 1 & 2 (cont'd.)
Safety Evaluation Summary: (cont'd.)
Analysis of the affected piping .and supports is in accordance with the requirements of Section IIX of the ASME Code 1974 Edition, and in compliance with the requirements specified in the, USAR.
Each valve weighed in excess of 1000 lbs.; therefore, when the valves were loaded into the reactor building and main steam tunnel they were treated as heavy loads. This required that the rigging arrangements and load paths chosen met the requirements of the Heavy Loads Study (USAR Appendix 9C) and NUREG-0612.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 29 of 93 Safety Evaluation No.: 91-094 Implementation Document No.: Simple Design Change SC2-0391-91 USAR Affected Pages: Section 9.4 System: Control Building HVAC Title of Change: Remote Shutdown Room Air Conditioner Filter Differential Pressure High Alarms Description of Change:
Originally, the instrument sensing lines for the remote shutdown room air conditioning units (2HVC*ACU3A/B) air filter differential pressure (D/P) switches (2HVC-PDIS66A/B) were installed such that the low side of the D/P switch sensed the filter discharge This while the high side of the D/P switch sensed configuration did not sense true filter D/P room pressure.
and resulted in frequent control room trouble alarms.
This change connected the instrument sensing lines for the high side of the D/P switch to duct work on the suction side of the air filter to allow true filter D/P to be sensed.
Safety Evaluation Summary:
The remote shutdown room 2HVC*ACU3A/B filter differential function.
pressure switches (2HVC-PDIS66A/B) serve no safety They are required to alert the main control room that the air filters should be changed out to maintain proper operation of 2HVC*ACU3A/B. Their design function is not altered by this change to the sensing line location.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page'30 of 93 Safety Evaluation No.: 91-097 Xmplementation Document No.: Temporary Mod.91-093 USAR Affected Pages: N/A System Makeup Water Treating (WTS)
Title of Change: Ecolochem Purge Bypass Line Description of Change:
The WTS system continues to employ the temporary trailer-contained demineralizer (Ecolochem) to supplement the existing system. This temporary modification installed a purge line (flexible hose) from the Ecolochem trailer to vent valve connection 2WTS-V263. The addition of this purge line provides a route for injection of the initial Ecolochem effluent (predemineralized water) directly into the waste neutralizing tank (2WTS-TK1), thus bypassing the existing waste sump and reducing the work load on the sump pumps.
Safety Evaluation Summary:
The temporary addition of the purge bypass hose will have no adverse impact on the WTS system operation or any other component or system required to effect or support safe shutdown of the reactor, or to perform in the operation of reactor safety features. Hose installation followed all safety precautions, and all work was in accordance with approved specifications and procedures. The flow to the makeup waste neutralizing tank will contain no radiological contaminated fluid.
Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 31 of 93 Safety Evaluation No.: 92-007 Implementation Document No.: EDC 2M10281 USAR Affected Pages: Section 4.6 Systems Control Rod Drive (RDS)
Title of Change! CRD Pump Suction Filters Normal Lineup Description of Change:
The original RDS system design intended for both CRD pump suction filters (2RDS-FLT1A, 1B) to be in service. This change modified filter is normally in the system configuration such that one service, with the other in standby. This change was accomplished by maintaining valve 2RDS-V3B normally closed.
Safety Evaluation Summary:
The revised system lineup allows the standby filter to be placed in service when a high filter differential pressure alarm is reached. Past experience and consultation with General Electric indicate that this lineup is acceptable and each filter is capable of handling full system flow.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page 32 of 93 Safety Evaluation Mo.s92-015, Implementation Document No.s Temporary Mod.92-024 USAR Affected Pages: N/A System: Fire Protection - Water (FPW)
Title of Changes Connecting Fire Protection System to Condenser Tube Flushing Rig Description of Changes This temporary modification involved the installation of a temporary fire hose from fire hose reel 2FPW-FHR14 for the flushing of the main condenser tubes during Refueling Outage 2.
The connection at the fire hose reel required the installation of an angle Y-gate with two valves; one to isolate the fire-fighting hose and the other to isolate the temporary fire hose used for tube flushing.
Safety Evaluation Summary:
This temporary modification will have no impact on the safe operation or capability to keep the plant in the safe shutdown condition.
Fire protection capabilities are not degraded at any time, because system pressure and flow will be maintained in the fire protection system while this temporary modification is in effect.
The makeup capabilities of the fire protection system far exceed the demands required by the tube flushing rig for the intermittent usage expected for the duration of flushing.
Based on the evaluation performed, it. is concluded that this change does not involve an unreviewed safety question and does not decrease the effectiveness of the fire protection program.
Safety Evaluation Summary Report Page 33 of 93 Safety Evaluation No.: 92-016, Rev. 1 Implementation Document No.: Procedure NIP-ECA-01 USAR Affected Pages: Sections 1.10, 13.4 System: N/A Title of Change: Operations Experience Assessment Description of Change:
This safety evaluation takes credit for the Deviation/Event Reporting (DER) process as an equivalent method for ensuring the proper dissemination and use of internal and external operating experience. The DER process eliminates the need for the OEA Committee.
Safety Evaluation Summary:
Revision 1 of this safety evaluation accounts for the fact that procedure NIP-SRE-01 was never issued. NIP-SRE-01 was referenced by Safety Evaluation 92-016 as containing the necessary procedural controls, in part, for'he processing of operating experience information. The discontinuation of NIP-SRE-01 is acceptable because the portions of NIP-SRE-01 that were taken credit for in the original safety evaluation were incorporated into the current revision of DER procedure NIP-ECA-01.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page 34 of 93 Safety Evaluation No.: 92-021 Implementation Document No.: Simple Design Change SC2-0247-91, EDC 2F00467A USAR Mfected Pages: Section 9.2 System: Circulating Water System (CWS), Makeup Water Treating System (WTS)
Title of Change: Circulating Water Pump 2CWS-P1A through P1F Seal Water Line Replacement Description of Change:
This change replaced the carbon steel circulating water pump seal water lines, including valves, with stainless steel material.
This change was necessary because the piping was experiencing reduced flow as a result of Microbiologically-Influenced Corrosion, iron oxides and sediment. To facilitate this change, a manual isolation valve was added to the makeup water seal water supply header.
Safety Evaluation Summary:
The circulating water and makeup treating systems are not required to effect or support safe shutdown of the reactor or to perform in the operation of reactor safety features. Changes made as described will enhance pump and system performance.
System function, operability, and integrity are not affected.
Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.
this
Safety Evaluation.
Summary Report Page 35 of 93 Safety Evaluation No.: 92-034 Implementation Document No.: DER 2-92-Q-0144 USAR Affected Pages: N/A System: Service Air (SAS), Condensate Demineralizer (CND)
Title of Change: Connecting the Service Air System to the Condensate Demineralizer System Description of Change:
This change involved the use of a mechanical jumper, i.e., red rubber hose, from the service air system (SAS) to the condensate demineralizer system (CND) in order to blow air into the piping on the low conductivity waste pumps, 2CND-P5A and P5B, if they became resin bound.
Safety Evaluation Summary:
The SAS and CND systems are not required to effect or support the safe shutdown of the reactor or to function in conjunction with any reactor safety features. The portions of the systems affected by this change are located in an area of the plant which does not require components to be seismically supported.
Connecting the two systems to unplug the low conductivity waste conductivity pumps does not degrade the design, function, or method by which the systems perform their design functions.
The hose(s) which are used for, the mechanical jumper are rated for a pressure greater than the design pressures of the affected SAS and CND piping to ensure that no failures would cause any radiological spills.
Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.
this
Safety Evaluation-Summary Report Page 36 of 93 Safety Evaluation No.: 92-036 Implementation Document No.: Simple Design Change SC2-0077-92 USAR Affected Pages: Sections 9.3, 11.4 System: Radioactive Solid Waste (WSS)
Title of Change: Radioactive Waste Dewatering System Description of Change:
This change added a radioactive waste dewatering system to reduce the volume of waste to be disposed. The system is known as RDS-1000, supplied by Chem Nuclear Systems, Inc. (CNSI). The equipment associated with RDS-1000 was located in the radwaste building truck bay. RDS-1000 is a self-contained, freestanding portable system for dewatering radioactive spent bead resins and filter sludge in a variety of liners to meet the current disposal criteria at low-level waste disposal facilities.
Safety Evaluation Summary:
The NRC has reviewed the design and operation of the RDS-1000 system, as described in CNSI Topical Report No.
RDS-25506-01-P/NP, Revision 1, and has concluded that the topical report is acceptable for referencing. NMPC reviewed the CNSI-supplied information for plant-specific impacts including design, testing, process control, ALARA considerations, floor loading, fire protection, plant services requirements, and waste handling and disposal. No adverse plant impacts were identified.
Based on the evaluation performed, does not involve an it unreviewed is concluded that this safety question.
change
Safety Evaluation Summary Report Page 37 of 93 Safety Evaluation No.: 92-038, Rev. 2 & 3 Implementation Document No.: Procedures GAP-OPS-01 Rev. 00, GAP-POL-Ol Rev. 01, GAP-RPP-01 Rev. 00, NEP-POL-300 Rev. 01, NIP-FPP-Ol Rev. 01 USAR Affected Pages: Sections 9A.3, 12.1, 12.5, 13 lg 13 2g 13 5
~ ~ ~
System: N/A Title of Change: Nine Mile Point Unit 2 Reorganization Description of Change:
Chapter 13 of the USAR describes the organization responsible for operation of Nine Mile Point Unit 2. This change addresses revisions to the Nuclear Division organizational structure.
Departments and positions were redefined and reorganized to enhance the flow of communication and productivity of the Nuclear Division. Affected areas of the Division organization include Executive, Generation and Quality Assurance.
Safety Evaluation Summary:
The organizational changes provide the Nuclear Division organization with resources to be both efficient and effective while meeting NRC guidance. No new functional areas or chains of command were created contrary to this guidance. The changes did not reduce the effectiveness of supervision or the ability of groups or individuals to perform activities necessary to ensure safe operation or shutdown of the plant. Positions specific to Unit 2 meet ANSI/ANS-3.1-1978 requirements as endorsed by Regulatory Guide 1.8. Positions with site-related responsibilities meet both ANSI/ANS-3.1-1978 and ANSI/N18.1-1971 as endorsed by Regulatory Guide 1.8.
Based on the evaluation performed, it is concluded that these safety question.
changes do not involve an unreviewed
Safety Evaluation Summary Report Page 38 of 93 Safety Evaluation No.: 92-040, Rev. 1 Implementation Document No.: Simple Design Change SC2-0119-92 USAR Affected Pages: N/A System: Neutron Monitoring Title of Change: SRM Count Rate Upscale Alarm Setpoint Change Description of Change:
This change revises the count rate upscale alarm of the source range monitors (SRMs) during core offload/reload. Specifically, the upscale alarm setpoint will be changed from 1 x 10~ cps to 100 cps (with an analytical limit of 200 cps). This change will be implemented only during refueling outages involving a complete core offload/reload, and allows the SRM count rate upscale alarm to fulfillthe continuous audible alarm function required by Technical Specification 3.9.2b.
Safety Evaluation Summary:
The SRM short period audible alarm had previously been considered as satisfying the continuous audible alarm requirement of Technical Specification 3.9.2b. However, during complete core offload/reload, when the reactor core is nearly empty, SRM short period alarms occur frequently even though the reactivity condition of the reactor is not changing. This represents a nuisance alarm.
Reducing the setpoint of the count rate upscale alarm provides the control room with an adequate indication of the core status, and provides an audible alarm to meet the intent of Technical Specification 3.9.2b. No change is made to the SRM channel trip unit outputs which perform trip functions. No transient or accident analyses take credit for any alarm from the SRM system.
Based on the'valuation performed, it safety change does not involve an unreviewed is concluded that question.
this
Safety Evaluation Summary Report Page 39 of 93 Safety Evaluation No.!92-041 Implementation Document No.: Mod. PN2Y92MX006 USAR Affected Pages: Section 9.2 System: Service Water Title of Change! Revise Logic for Service Water Valves MOV95A/B and MOV66A/B Description of Change:
J This modification was initiated to prevent disabling the Division III diesel due to service water pressure perturbations from sequential single divisional loss of power or complete LOOP after the Division II diesel was running. This change installed a new time delay relay for the supply header (SWP) transmitter allowing approximately 78 seconds prior to closing these valves on loss of header pressure. The existing time delay in the above relay was removed, thus preventing the loss of the HPCS diesel on any low pressure header signal.
In addition to these changes, the existing time delay settings for the Division I and II SWP discharge valves was increased to delay closure of the discharge valves on low SWP header pressure.
This delay allows time for the service water supply valves to the Division II diesel to close prior to the discharge valves in the event of low service water header pressure. This allows time for the service water header to repressurize in the event of a line break downstream of the Division III service water supply valves.
Safety Evaluation Summary:
This modification allows the service water system the required time to repressurize the supply header in response to a LOOP, prior to isolating, the HPCS diesel cooling water supply. In the event of a l'ine; rupture, in the service water piping downstream of 2SWP*MOV95A/B',-
piping and the it -'also allows time for the isolation of that time required to repressurize the supply header to the Division I and II diesel prior to isolating the cooling water supply to'hose diesels.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 40 of 93 Safety Evaluation No.: 92-042 Implementation Document No.: EDC 2M00365 USAR Affected Pages: Sections 9.2, 9.3, 11.2 System: Service Water, Radioactive Liquid Waste and Instrument Air Title of Change: Revise the Position of a Series of Valves on Figures to Agree with the Operating Procedures Description of Change:
The following valves were diagrammatically shown incorrectly in the USAR figures and corresponding system P&IDs:
Service Water System: 2SWP-V933, 2SWP-V934 Radioactive Liquid Waste System: 2LWS V83 I V276 I V28 1 g V306 g V328 ~
V329 g V330 g V33 1 g V359 g V360 g V361 g V362 g V374 g V375 g V378 g V379 g V557 g V558, V570 and V571 Instrument Air System: 2IAS-V176, 2IAS-V177 The service water system valves have been revised from normally closed to normally open. The radioactive liquid waste system valves have been revised from normally open to normally closed.
The instrument air system valves have been revised from normally closed. to locked closed. These changes were made to reflect actual valve configurations and to agree with the associated operating procedures.
Safety Evaluation Summary:
The changes to system PEIDs and corresponding USAR figures reflect the position of various valves in performing their intended design function. The revised drawings agree with design requirements, normal plant configuration and operating procedures.
Based on the evaluation performed, it is concluded that this documentation change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 41 of 93 Safety Evaluation No.: 92-044 Implementation Document No.: DER 2-92-Q-1740, EDC 2F00572 USAR Affected Pages: Section 9.5 System: Auxiliary Boiler System (ABM)
Title of Change: Auxiliary Boiler Conductivity Specification Changes Description of Change:
This change updated Hydro Steam Industries (HSI) vendor manual N2H32800HTEXCH001 (N20843) and the USAR to reflect the current conductivity range of auxiliary boiler (2ABM-B1A and 2ABM-BlB) water at low steam demands. The boilers must operate at levels below those previously referenced in the above documents due to low steam demand.
Safety Evaluation Summary:
This change revised the acceptable conductivity operating range to reflect current boiler operation. No changes were made which would impact any safety-related components, systems or structures required for safe operation or required to maintain the plant in a safe shutdown condition.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 42 of 93 Safety Evaluation No.!92-045, Rev. 0 & 1 Implementation Document No.! Procedures ARC-CSA2, ARC-CS02 USAR Affected Pages: N/A System! N/A Title of Change: Acceptability of >>CAVSPAN>>
Gantry/Manlift at NMP2 Description of Change!
This temporary change involves the use of the >>CAVSPAN>> system for outage decontamination activities.
The >>CAVSPAN>> is a reactor refueling cavity-spanning gantryg designed by Applied Radiological Control, Inc. (ARC), and used for suspending work cages in support of decontaminating the reactor cavity and internals storage pool. The >>CAVSPAN>> has two manlifts attached to the beam and two gantries with rail wheels which travel along the refueling bridge rails. Movement of the unit is done manually. The maximum load which is allowed to be suspended in either of the two work cages is 800 pounds including personnel weight and equipment.
Safety Evaluation Summary:
The use of the >>CAVSPAN>> system for the decontamination of the reactor cavity and internals storage pool is with the plant in cold shutdown condition after the reactor vessel head is in place and prior to plant restart. Stress analyses of the beam, manlift, and drive assembly have been performed to demonstrate adequate safety factors, and a load test of the structure for 125 percent of rated load was also conducted. Appropriate criteria from USAR Appendix 9C, "Control of Heavy Loads," were also applied.
During a seismic event, if the CAVSPAN was to jump from the refueling bridge rails and fall into the reactor cavity or the reactor internals storage pool, no systems, equipment or components required to maintain the safe shutdown of the plant would be damaged.
h f Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 43 of 93 Safety Evaluation No.: 92-046, Rev. 0 & 1 1mplementation Document No.: Major Order No. 1644 USAR Affected Pages: Sections 1.2, 2.4 Systems N/A Title of Change: Nine Mile Point Compressed Bottled Gas Storage Facility Description of Change:
This change consists of the construction of a new bottled gas storage facility. The new storage facility is a nonsafety-related structure and is located outside the protected area south of the Unit 2 warehouse. The area of the new facility is about 2500 sq. ft. with interior ceiling height about 15 ft.,
and is designed to accommodate 550 bottles of various compressed gases.
The facility consists of two areasi the east area is designated is for storage of the flammable bottles, and the west area designed for storage of the nonflammable bottles.
Safety Evaluation Summary:
The construction of the new storage facility does not disturb those attributes of the site in the immediate vicinity of the plant, which safely divert. the local probable maximum precipitation (PMP) runoff overland to Lake Ontario. Also, since the new facility is low in elevation and outside the protected area, this location will not create any wind disturbances which may affect the atmospheric dispersion factor study.
The effects of an accidental nitrogen gas release from the facility on control room habitability were evaluated. The
. potential for missiles as a result of fire or explosion was also considered; No adverse impacts were identified.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 44 of 93 Safety Evaluation No.: 92-049 Implementation Document No.: Procedure N2-OP-55 USAR Affected Pages: Section 9.4 System: Turbine Building Ventilation Title of Change: Alternate Operation of the Turbine Building Ventilation System Description of Change:
This change permits various fan configurations and damper controls on the turbine building ventilation system to provide additional capability to maintain the building at subatmospheric pressure. The change allows manual manipulation of dampers HVT-AOD101 & 102, and/or operation with one supply fan and two exhaust fans to ensure the ability to maintain slight negative pressure in the turbine building.
Safety Evaluation Summary:
The change to turbine building ventilation system opera'tion enhances the ability to maintain slight negative pressure in the turbine building with respect to atmospheric, thus maintaining the design bases of the turbine building as described in the USAR. This is a nonsafety-related system, and this change has no impact on safe operation or shutdown of the plant. This change will not result in any additional radioactive releases from the plant and, in fact, will enhance the ability to prevent an unmonitored release from the plant. No equipment has been added, deleted or modified as a result of this change.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 45 of 93 Safety Evaluation No.: 92-050, Rev. 1 Implementation Document No.: Temporary Mods.92-053 and 92-055 USAR Affected Pages! N/A System: Vital Bus (VBB)
Title of Change: Temporary Bypass of 2VBB-UPS1C and 2VBB-UPS1D Description of Change:
These temporary modifications connected the loads of uninterruptible power supplies (UPSs) 2VBB-UPS1C and 2VBB-UPS1D the individual UPS directly to the alternate power sources of(2NJS-US5 units, which are the 600-V ac stub buses and 2NJS-US6).
The purpose of this change was to permit replacement of the subject UPS units under Modification PN2Y89MX042, while maintaining operable the plant communications and essential lighting loads that are supplied by the UPS atunits. oneOnly one of time.
these temporary modifications was in place any Safety Evaluation Summary:
The loads of the affected UPS units (2VBB-UPS1C and 2VBB-UPS1D) are essential lighting and communication circuits. These temporary changes eliminated the 125-V dc battery backup by connecting the UPS loads directly with the'alternate power supplies (the 600-V ac stub buses 2NJS-US5 and 2NJS-US6) which are normally fed from the station transformer 2STX-XNS1. In the event of a loss of normal power, these stub buses would be connected to offsite power through the reserve station transformers 2RTX-XSR1A/B. In the event of loss of normal and offsite power, these stub buses could be manually connected to the emergency diesel generators (except in a LOCA situation).
The emergency and the 8-hour battery-pack lighting systems which are required for the safe shutdown equipment areas of the plant, including the access and egress routes thereto (required per 10CFR50 Appendix R), are not affected by these temporary modifications. Alternate methods of communication are also available during the time that the page party/public address system is without the uninterruptible power source.
Based on the evaluation performed, not involve an it is concluded that these unreviewed safety question.
temporary changes do
Safety Evaluation Summary Report Page 46 of 93 Safety Evaluati.on No.: 92-052 Xmplementation Document No.: Temporary Mod.92-044 USAR Affected Pages: N/A System: N/A Title of Change: Reroute the Security Fence to Support the Demolition of "Area Complex" Building and the Construction of the Swing Building Description of Change:
This temporary modification installed a "bubble fence" and rerouted the security fence to segregate the area complex site from the protected area of Nine Mile Point so that activities associated with construction of the swing building would be outside the security zone.
Safety Evaluation Summary:
This temporary modification does not result in a significant elevation change in the flooding levels within the berm area of the Nine Mile Point site. Therefore, rerouting the security fence and construction of the "bubble fence" do not disturb those attributes of the site in the immediate vicinity of the plant which safely divert the local probable maximum precipation (PMP) runoff overland to Lake Ontario.
The temporary "bubble fence" was constructed following security procedures and regulation. The fence is equipped with security intrusion detection, and a closed-circuit TV camera (CCTV) was installed in accordance with 10CFR Part 73.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 47 of 93 Safety Evaluation No.: 92-056 Implementation Document No.: Technical Test Procedure N2-TTP-LTC-9001 USAR Affected Pages: N/A System: Normal and Reserve Station Service Transformers Title of Change: Setup and Testing of Load Tap Changers Description of Change:
This change consisted of testing to verify setup of the load tap changers (LTC) automatic controls for both reserve station service (RSS) transformers and the normal station service (NSS) transformer, and to functionally verify the LTC's ability to control bus voltage when operated in automatic. The controls of the LTCs had been left in the manual mode of operation since the startup of the plant. The tests were performed during normal plant operation.
Safety Evaluation Summary:
Both the RSS transformers (2RTX-XSR1A and 1B) and the NSS transformer (2STX-XNS1A) are nonsafety related. Development and performance of the LTC tests considered the following:
- 1. The test is performed one transformer at a time.
- 2. The test follows the vendor-recommended alternate method of testing.
- 3. All precautions and prerequisites of the test procedure are strictly adhered to.
- 4. Loss of voltage/sustained degraded voltage will start the standby diesel generator and carry the emergency loads.
- 5. An overvoltage condition is monitored and corrective action taken as required by the test procedure.
Based on the evaluation performed, it is concluded that this test does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 48 of 93 Safety Evaluation No.: 92-057, Rev. 1 Implementation Document No.:. Simple Design Change SC2-0018-92 USAR Affected Pages! Section 5.4 System: Reactor Core Isolation Cooling (RCIC)
Title of Change: RCIC Turbine Exhaust Pressure Trip Setpoint Modification Description of Change:
This modification increased the nominal RCIC turbine exhaust pressure trip setpoint from 25 psig to 50 psig. This change provides a longer period of RCIC operation before its turbine is tripped off by high exhaust pressure due to increased primary containment pressure following postulated Station Blackout (SBO) events.
Safety Evaluation Summary:
Operation of Unit 2 with the increased RCIC turbine exhaust pressure trip setpoint does not affect any of the USAR accident and/or transient analyses. The new exhaust pressure trip setpoint of 50 psig is well within the design rating of 165 psig for the RCIC turbine exhaust casing and the design rating of 150 psig, for the RCIC exhaust piping. The new setpoint will not have an adverse impact on the RCIC system pumping performance as discussed in Section 4 of GE Report NEDE-22017.
This modification will result in an increased RCIC turbine gland seal leakage rate. The radiological effects of this higher leakage rate will not exceed the limits of 10CFR20.
Evaluation of equipment qualification data concludes that the RCIC system will continue to operate at the elevated RCIC room temperatures resulting from the gland seal leakage.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 49 of 93 Safety Evaluation No.: 92-058 Implementation Document No.: Temporary Mods.92-071, 92-072 USAR Affected Pages: N/A System: Primary Containment Purge (CPS)
Title of Change: Temporary Replacement of 2CPS*AOV111 Description of Change:
These temporary modifications removed suppression chamber purge for repairs and line containment isolation valve 2CPS*AOV1112CPS*AOV111 replaced
- repaired, ititwith was two blank reinstalled, flanges.
tested, After and returned to was service.
Safety Evaluation Summary:
These temporary modifications allow 2CPS*AOV111 to be removed for repairs and at the same time provide containment isolation in accordance with Technical Specification 3.6.3. The blank flange meets or exceeds the design pressure, temperature, and leakage requirements that were met by 2CPS*AOV111.
Installation of the blank flanges results in the unavailability .
of the normal suppression chamber purge flow path. In the event of an accident occurring while the blank flanges are installed, alternate pathways are available to control hydrogen concentration in the suppression chamber, perform emergency containment venting, and conduct postaccident cleanup operations.
Based on the evaluation performed, does not involve it an is concluded that this unreviewed safety question.
temporary change
Safety Evaluation Summary Report Page 50 of 93 Safety Evaluation No.: 92-059, Rev. 1 Implementation Document No.: Unit 1 Mod. No. N1-86-026 USAR Affected Pages: N/A Systems TSC Emergency Ventilation (System f212)
Title of Change: Addition to TSC Ventilation Control Panel Description of Change:
This modification installed indication lights for Technical Support Center (TSC) emergency ventilation fan FN-1 and dampers 212-42, 212-31 and 212-87, located in the charcoal filter room, as well as indication lights for normal/emergency power source.
A digital timer was also wired to key switch KS-2. The timer runs only when the TSC ventilation system is in the emergency mode of operation. All indication is provided on the TSC ventilation control panel ATPC-1, located in the TSC.
Safety Evaluation Summary:
This modification provides indication at the TSC ventilation control panel to monitor status of ventilation equipment related to the TSC, without requiring occupants to exit the TSC.
Allowing personnel to remain in the TSC during emergency conditions decreases the possibility of personnel contamination and loss of valuable TSC'personnel time.
The new equipment/material introduced by this modification only affects the TSC emergency ventilation system. It is isolated from other areas of NMP1 and NMP2 which could affect the safe shutdown of either plant.
Based on the evaluation performed, it is concluded that this change does. not involve an unreviewed safety question.
Safety. Evaluation Summary Report Page 51 of 93 Safety Evaluation No.: 92-062 Implementation Document No.: Simple. Design Change SC2-168-92 USAR Affected Pages: Section 9.2 System: Service Water (SWP), Reactor, Building Ventilation (HVR)
Title of Change: Simple Design Change to Enhance Unit Cooler Testing Related to Secondary Containment Drawdown Description of Change:
This change added flanges to the service water pipe lines to unit coolers 2HVR*UC404B and 2HVR*UC407E to facilitate performance testing of this equip'ment. Testing is performed to ensure unit, cooler operability.
Safety Evaluation Summary:
This change enhances the ability to performance test unit coolers 2HVR*UC404B and 2HVR*UC407E. The original design basis of the piping supports is maintained. Piping installation and material is ASME III Class 3 under the jurisdiction of ASME XI. Thus, the modification will not adversely affect the safety function of any of the structures, systems or components.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 52 of 93 Safety Evaluation No.: 92-063 Implementation Document No.: Simple Design Change SC2-281-92 USAR Affected Pages: Section 9.3 System! Instrument Air System Title of Change: Revise Updated Safety'Analysis Report Instrument Air/Nitrogen Particle Size Limit from 3 Microns to 40 Microns Description of Change:
As a result of Generic Letter 88-14, the particle size limit for instrument air/nitrogen has been revised from 3 microns to 40 microns. The 3-micron particle size limit was based on ANSI MC11.1-1976, "Quality Standards for Instrument Air" (now Standard ISA-S7.3). The new 40-micron limit is based upon actual vendor requirements stated in the vendor instruction manuals to protect equipment from being damaged by particulate intrusion.
Safety Evaluation Summary:
The 40-micron limit established by the equipment vendors will avoid plugging and wear/erosion of air passages and orifices of equipment due to particulate intrusion. This change took exception to*the 3-micron limit contained in Standard ISA-S7.3 and established a new 40-micron limit based on vendor recommendations. All plant instrument air and nitrogen users will function designed with the new particle size limit of 40 as microns with no adverse effects.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page 53 of 93 Safety Evaluati.on No.: 92-064, Rev. 1 Implementation Document No.: LDCN U-1674 USAR Affected Pages: Section 3A.34 System: N/A Title of Change: Addition of IMAGES Computer Software to USAR Appendix 3A for Structural/Equipment Analysis Description of Change:
This change permits the use of IMAGES computer software in qualifying safety-related structures and equipment, as an alternative to other well-known mainframe computer programs.
IMAGES is a complete desktop Finite Elements Analysis Package for the PC (by Celestial Software, Inc.), with the capability to perform static, thermal, modal and dynamic analyses in ways similar to STRUDL, STARDYNE, and ANSYS mainframe computer programs.
Safety Evaluation Summary:
The IMAGES computer software has been benchmarked to the analytical results published in the literature in accordance 3.9.
with CSI's Quality Assurance Program and Standard Review Plan This QA Program has been audited and accepted by NMPC QA. It meets the intent of NMPC's Nuclear Division Software QA Program Plan. In addition, NMPC performed limited program verification on IMAGES by satisfactorily comparing its solutions to solutions obtained from the recognized, widely-known STRUDL program. The IMAGES program meets the methodologies and commitments described in USAR Sections 3.7, 3.9, 3.10 and Appendix 6A.
Based on the evaluation performed, does not involve an it unreviewed is concluded that this safety question.
change
Safety Evaluation Summary Report Page 54 of 93 Safety Evaluation No.: 92-066 Implementation Document No.: Procedure N2-OP-10A USAR Affected Pages: N/A System: Circulating Water (CWS)
Title of Change: Defeat Low Suction Pressure Trip Interlock for CWS Pumps Description of Change:
This temporary change defeats the low suction pressure trip for the operating CWS pumps while starting a standby pump. Leads are lifted at the CWS pump switchgear to defeat this trip function, while at the same time maintaining the low suction pressure alarm and computer point operable. Following successful pump start, the lifted leads are relanded to restore the system to its design configuration.
Safety Evaluation Summary:
The purpose of this change is to prevent inadvertent operating pump trip from a momentary low suction pressure condition while starting a standby pump. Sufficient instrumentation will remain available to alert the Operators should a sustained degraded condition exist. This activity is procedurally controlled and requires independent verification for lifting and relanding the leads. The low suction pressure trip, as well as the entire CWS system, does not perform a safety-related function.
Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 55 of 93 Safety Evaluation No.: 92-067, Rev. 1 Implementation Document No.: Simple Design Change SC2-0253-92 USAR Affected Pages: Section 9.2 System: Service Water Title of Change: SDCg SC2-0253-92, Delete Service Water to Circulating Water Makeup Header Pressure Low Annunciator (Nuisance Alarm)
Description of Change:
This simple design change deleted control room annunciators CEC601123 and CEC601213. Pressure switches 2SWP*PSL142A, 2SWP*PSLX142A, 2SWP*PSL142B, 2SWP*PSLX142B and their associated alarm circuits were disconnected and spared in place.
2SWP*PT142A and 2SWP*PT142B continue to provide service water to circulating water makeup header pressure signals available for information only read-out on the PMS computer display in the control room.
Safety Evaluation Summary:
This simple design change eliminated control room annunciators which are continuously in alarm and a nuisance and distraction to the control room operators. This change has no impact on the safe operation or shutdown of the plant. Adequate indication exists without these annunciators, to keep, operators informed as to the status of the service water system to circulating water system makeup headers.
Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.
this
Safety Evaluation Summary Report Page 56 of 93 Safety Evaluation No.!92-070 Implementation Document No.: N/A USAR Affected Pages: Sections 9A.2, 9A.3, 13.1 System: N/A Title of Change: Reduction Fire Brigade Staffing through Partial Combination of the Unit Fire Brigades Description of Change:
This change reduces the unit Fire Brigade staffing to a minimum of a Fire Chief and two Fire Fighters. This results in a minimum site response organization of five Brigade members.
Safety Evaluation Summary:
Establishing a unit staff size of a Fire Chief and two Fire Fighters achieves the requirements of 10CFR50 Appendix R and BTP CMEB 9.5-1,'hich requires that at least five Brigade members respond to a fire. Of these five responders, the Fire Chief and two members must be familiar with the effects of fire and fire suppression activities on plant systems. The reduction in unit-dedicated Fire Brigade staffing levels will nowt result in a lesser response to a fire (either in number of personnel fighting the fire or in a significant increase in their response time), or in a loss of fire watch or surveillance/maintenance activities.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question and does not decrease the effectiveness of the fire protection program.
Safety Evaluation Summary Report Page 57 of 93 Safety Evaluation No.: 92-071 Implementation Document No.: Procedure N2-OP-1 USAR Affected Pages: N/A Systems Main Steam (MSS), Reactor Core Isolation Cooling (RCIC)
Title of Change: Defeat Interlock of MSS Drain Valve 2MSS*MOV189 Description of Change:
Valve 2MSS*MOV189 is a drain valve from the RCIC steam supply line that connects into a common drain line for the main steam lines. The valve has an interlock that maintains it open whenever one of the RCIC isolation valves is shut, to remove any condensation from the RCIC steam supply header. This temporary alteration defeats this interlock to allow valve closure when a main steam line (MSL) is isolated.
Defeating the 2MSS*MOV189 interlock is accomplished by lifting the lead on terminal point. AA-7, 2CEC-PNL856 Bay D. This removes the automatic open function only.'emote manual control of the valve and the auto close function is unaffected.
Safety Evaluation Summary:
With a MSL isolated, a drain path is established to allow steam/condensate in the main steam piping to drain to the main condenser, as recommended in General Electric Service Information Letter Number 404. If the RCIC system becomes isolated due to closure of valves 2ICS*MOV121 or 2ICS*MOV128, 2MSS*MOV189 will automatically open to establish a drain path. The subsequent resultant pressure in the combined MSL drain header could affect the proper drainage of the main steam lines.
The defeating of the interlock for 2MSS*MOV189 is only required when RCIC is already isolated and considered inoperable.
Technical Specification 3/4.7.4 already directs actions to be taken in the case of an inoperable RCIC system.
Control of defeating and restoring this interlock will be in accordance with approved procedures.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page 58 of 93 Safety Evaluation No.: 92-072, Rev. 1 Implementation Document No.: Special Operating Procedures N2-SOP-01, 02, 03 USAR Affected Pages: N/A System: N/A Title of Change: Station Blackout (Complete Loss of AC Power)
Description of Change:
Three special operating procedures (SOPs) have been developed for coping with a station blackout (SBO). These SOPs may be executed concurrently with the applicable emergency operating procedures (EOPs) ~
1~ N2-SOP-01 is a flowchart-based procedure designed to guide the operators through the actions required to cope with the SBO event as well as to give direction for recovery.
- 2. N2-SOP-02 is a support procedure for N2-SOP-01, and provides specific directions for operators in response to a SBO event to achieve stable shutdown conditions.
3 ~ N2-SOP-03 was created in order to consolidate the actions needed for power restoration into a single procedure. Although N2-SOP-03 contains some SBO specific steps (i.e., can only be performed during an SBO), it can also be used for, any loss of offsite ac power or loss of onsite emergency ac power (i.e., non-SBO power losses).
Safety Evaluation Summary:
The results of the SBO analysis show that with certain specific operator actions, NMP2 can operate longer than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without jeopardizing core cooling or the containment capability.
Operator actions needed to achieve this capability include shedding of nonessential battery loads, maintaining manual RCIC flow control, and the bypassing of certain RCIC isolation logics.
These actions are addressed in N2-SOP-01, 02, and 03, which direct operators to make use of available plant systems to cope with a SBO and do. not conflict in any way with established EOPs.
Safety Evaluation Summary Report Page 59 of 93 Safety Evaluation No.: 92-072, Rev. 1 (cont'd.)
Safety Evaluation Summary: (cont'd.)
The actions'pecified in N2-SOP-01, 02 and 03 are in accordance with commitments made by NMPC to the NRC. NRC acceptance of NMPC's responses and commitments regarding the SBO issue were documented in the NRC safety evaluation dated May 29, 1991, and supplements dated November 21, 1991, and February 7, 1992.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page 60 of 93 Safety Evaluation No.: 92-074, Rev. 0, 1 & 2 Implementation Document No.: Simple Design Change SC2-0387-91 USAR Affected Pages: Section 9.3 System: Instrument Air (IAS)
Title of Change: IAS Strainer 40 Blowdown Valve Replacement Description of Change:
This simple design change replaced the blowdown connection from strainer 2IAS-STR40, including globe valve 2IAS-V355. The weight of the assembly, in addition to the force required to open and close the valve, had proven too great for the 1/4" NPT connection at the bottom of the strainer body. The replacement connection utilized tube products and compression fittings, and the globe valve was replaced with a NUPRO plug valve. This material is lighter, and operation of the blowdown isolation valve requires less torque.
Safety Evaluation Summary:
This simple design change will have no impact on the safe operation or safe shutdown capability of the plant.
The design of the replacement connection exceeds that of the existing connection and provides a leak-tight, easily maintained and operated strainer blowdown connection, thus maintaining system integrity.
Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.
this
Safety Evaluation Summary Report Page 61 of 93 Safety Evaluation No.: 92-076, Rev. 0, 1 & 2 Implementation Document No.: Mod. PN2Y89MX133 USAR Affected Pages! Section 9.2 System: Service Water (SWP), Reactor Building Ventilation (HVR),
Yard Structures Ventilation (HVY), Control Building A/C (HVC), Diesel Generator Bui:lding Ventilation (HVP)
Title of Change! Provide Breakout Spools for Unit Coolers not Related'to Drawdown Description of Change:
NRC Generic Letter 89-13 requires that safety-related components cooled by service water be adequately maintained and tested.
Currently at NMP2, the capability does not exist to isolate and test some safety-related unit coolers.
The scope of the modification includes provision of breakout spools to the supply and/or return lines of safety-related unit coolers that are not related to secondary containment drawdown.
Revision 1 to the safety evaluation was issued to indicate changes in the location of central flanges. Revision 2 to the safety evaluation was issued to revise the response to question D on the Certificate of Compliance to NRC standards.
Safety Evaluation Summary:
The modified piping is safety-related and the requirements of 10CFR50 Appendix B apply. The original design basis of the piping and supports is maintained. Piping installation and material The is ASME modification III will Class not 3 under adversely the jurisdiction of ASME XI.
affect the safety function of any of the structures, systems or components. In addition, the capability to safely shut down the plant and maintain safe shutdown condition will not be adversely affected. The it in a modification does not affect the Plant Technical Specifications, and components added or modified by this modification are not subject to any Technical Specification surveillances.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 62 of 93 Safety Evaluation No.: 92-077 Implementation Document No.: Procedure N2-CTP-CWS-807 USAR Affected Pages: N/A Systems Circulating Water System (CWS)
Water Treatment Hypochlorite (WTH)
Title oi Change: Justification of Circulating Water Sodium Hypochlorite Addition Description of Change:
This change allows for a procedurally-controlled temporary alteration to be used in lieu of the permanent plant WTH system to accomplish hypochlorite injection into the CWS system.
Safety Evaluation Summary:
The addition of sodium hypochlorite to the circulating water, by direct pumping from barrels into the cooling tower flumes, allows continued control of biological growth within the circulating water and the cooling tower while the WTH system is inoperative.
Securing the circulating water blowdown allows site Chemistry to achieve chlorine levels in the circulating water below limits set by the SPDES permit prior to reopening blowdown, or to enact proper controls as delineated within site procedures. The WTH system is not required for safe operation or shutdown of the plant. The method of adding sodium hypochlorite to the circulating water does not impact continued operation of the CWS system. This method of adding sodium hypochlorite to circulating water provides an alternate means of performing a nonsafety-related function that has no negative impact on the safe operation or shutdown of the plant.
Based on the evaluation performed, change does not involve an it unreviewed is concluded that this safety question.
Safety Evaluation Summary Report Page 63 of 93 Safety Evaluation No.: 92-079 Implementation Document No.: Simple Design Change SC2-282-92 USAR Affected Pages: Section 9.2 System: Service Water (SWP) & Reactor Building Ventilation (HVR)
Title of Change: Simple Design Change to Enhance Unit Cooler Testing Related to Secondary Containment Drawdown Description of Change:
This change added unions to the SWP pipe lines to unit cooler 2HVR*UC406 to facilitate flushing and performance testing of this equipment. Testing is required to ensure unit cooler operability.'afety Evaluation Summary:
This change enhanced the ability to flush and performance test unit cooler 2HVR*UC406. The capability to counteract various heat loads is dependent on the performance of the unit coolers.
The modification will not adversely affect the safety function of any of the structures, systems or components. The original design basis of the piping and pipe supports will be maintained.
The modification will not adversely impact safe shutdown of the plant or the capability to maintain the plant in a safe shutdown condition.
Based on the evaluation performed, change does not involve an it unreviewed is concluded that this safety question.
Safety Evaluation Summary Report Page 64 of 93 Safety Evaluation No.s92-080, .Rev. 0 & 1 Implementation Document No.s Calculation EC-032 USAR Affected Pages: Section 8.3 Systems N/A Title of Changes Update Diesel Loading Tables Description of Change:
Tables 8.3-1, 8.3-2 and 8.3-3 of the USAR tabulate the design basis accident diesel generator loading possibilities for the Division I, II and III diesel generators. Tables 8.3-5 and 8.3-6 are the totals of these loading tables. The basis for these tables is calculation EC-032. Table 8.3-4 lists the safety-related loads by power source; the bases for these numbers were test reports provided by vendors during plant construction.
The following changes were made:
1 ~ All the loads listed in the tables were reviewed against the GENE test data sheets for the motors and corrections were made as needed. Any corrections in loading in Tables 8;3-1 and 8.3-2 were also reflected in the total loading Tables 8.3-5 and 8.3-6.
- 2. The power factors for the 4-kV motors were revised to provide consistency between the various electrical analyses.
- 3. This change displays some of the data in Tables 8.3-1, 8.3-2 and 8.3-3 in a different format but all of the same basic information is still available from the tables.
4 Corrections to Table 8.3-3 were made as not all the loads powered by the Div. III diesel generator were originally
~
listed.
- 5. It is currently (MOV) assumed that 204 of motor-operated valve running loads are on after the initial 2 minutes of diesel generator loading sequence. This assumption is changed to assume 104 of the running loads are operating.
- 6. Revisions to Tables 8.3-5 and 8.3-6 were made to address administratively controlling the loading of the diesel generators.
Tables 8.3-1 and 8.3-2 show all the same loads as before; only the format for presenting this information has changed. Table 8.3-3 includes some existing loads that were not previously included in the load tabulations.
Safety Evaluation Summary Report Page 65 of 93 Safety Evaluation No.: 92-080, Rev. 0 & 1 (cont'd.)
Safety Evaluation Summary:
The layout of the table information has been changed to provide more information. This data has also been checked against the as-built data and updated where necessary.
The total kW load on the diesel generators is shown to have changed. The total kW load for all three diesel generators is still under the continuous rating of 4400 kW for the Div. I and II diesel generators and 2600 kW for the Div. III diesel generator.
It is now assumed that 10% of the MOVs are running after the first 2 minutes of diesel loading, which is still conservative.
A note for Tables 8.3-5 and 8.3-6 was changed to allow the use of NMP2 operating procedures to provide direction for controlling loading of the diesel generators during a loss-of-offsite power with delayed loss-of-coolant accident.
Based on the evaluation performed, changes do not involve an unreviewed it is concluded that these safety question.
Safety Evaluation Summary Report Page 66 of 93 Safety Evaluation No.: 92-081 1mplementatkon Document No.: Simple Design Change SC2-0289-091 USAR Affected Pages: Section 9.5 System: Engine-Driven Fire Pump Fuel Oi:1 Title of Change: Diesel Fire Pump Valve Description of Change:
This modification permanently replaced the spring return ball valve on the fuel oil supply piping to the diesel fire pump with a manual plug valve. The use of the manual valve in the locked open position complies with NFPA 20 and is in accordance with the regulatory guidance of Branch Technical Position (BTP) 9.5-1.
Safety Evaluation Summary:
This change assures reliable operation of the diesel fire pump by eliminating the potential for closing the fuel oil supply 20valve during a fire emergency. This change complies with NFPA and BTP 9.5-1 requirements. The ability to safely operate, or shut down the plant in the event of a fire, has not been adversely affected by this change.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 67 of 93 Safety Evaluation No.: 92-083, Rev. 1 Implementation Document No.: Procedures N2-OP-13, N2-OP-14 USAR Affected Pages: N/A System: Reactor Building Closed Loop Cooling (CCP), Turbine Building Closed Loop Cooling (CCS), Makeup Water (MWS)
Title of Change: Temporary Hose Installations on the Makeup Water System to Freeze Seal Activities 'upport Description of Change:
Freeze seals were installed to isolate a small portion of the makeup water system to allow maintenance on leaking flow totalizer indicator switches 2MWS-FQIS121 and 2MWS-FQIS122. In order to maintain the CCP and CCS systems operable while the freeze seals were established, hoses were routed to bypass the freeze seal location, allowing makeup water flow to surge tanks 2CCP-TK1 and 2CCS-TK1 as demand necessitated. The hoses were connected to either vent or drain'onnections on each side of the freeze seals.
Safety Evaluation Summary:
The temporary hose installations and freeze seals will have no impact on the safe operation or safe shutdown capability of the plant. Failure of a temporary hose or piping within the freeze area will not affect nuclear safety. Essential equipment receives cooling water from the safety-related service water system. Use of freeze seals and the temporary hoses is procedurally controlled.
Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 68 of 93 Safety Evaluation No.: 92-085 Implementation Document No.: Simple Design Change SC2-0335-92 USAR Affected Pages: Section 9.5 System: Miscellaneous Equipment and Floor Drains (DFM)
Title of Change: Auxiliary Boiler Building Equipment Drains Description of Change:
This evaluation addressed two issues concerning piping and drains associated with the auxiliary boiler building. The first issue was a documentation-only change to show small bore feedwater pump drain line 2-ABF-001-83-4 discharging to equipment drain 2DFM-ED2002 instead of 2DFM-ED2003. The second issue involved the installation of an 8" x 4" concentric pipe reducer to convert equipment drain 2DFM-ED2003 to a funnel equipment drain. The documentation change was made to accurately reflect existing plant conditions. The conversion of equipment drain 2DFM-ED2003 to a funnel drain increased the drain opening size so as to prevent water spillage and potential slab contamination.
Safety Evaluation Summary:
The documentation change provides clarification of the actual plant configuration and does not represent any physical change to the plant. The modification to drain 2DFM-ED2003 will enhance the operation and function of the drain and reduce the likelihood of water spillage and possible slab contamination. The original equipment drain design is not affected by these changes as no new water sources for the drain are being introduced.
Based on the evaluation performed, changes do not involve an unreviewed it is concluded that these safety question.
Safety Evaluation Summary Report Page 69 of 93 Safety Evaluation No.: 92-086 Implementation Document No.: Procedures NTP-TQS-201 Rev. 1, NTP-TQS-505 Rev. 0 USAR Affected Pages: Section 13.2 System: Nuclear Training Procedures (NTP)
Title of Change: Radwaste Operator Training Supervisor Change Description of change:
This change allows Radwaste Operator Training to be under the supervision of either the General Supervisoi Operations Training or the General Supervisor Technical Training. This option allows flexibility of program administration in the most efficient and cost-effective manner.
Safety Evaluation Summary:
This change improves distribution of administrative work load for Nuclear Training by allowing the General Supervisor Technical Training to supervise the Radwaste Operator Training program, while not precluding the General Supervisor Operations Training from supervising the program in the future. This change is strictly administrative and will not affect conduct of Radwaste Operator training. Training will continue to meet INPO accredited training program standards. Radwaste Operators are not Licensed Reactor Operators and, as such, the INPO criteria for an accredited training program are the same as for the rest of the Technical Training groups.
Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.
this
Safety Evaluation Summary Report Page 70 of 93 Safety Evaluation No.: 92-087 Implementation Document No.: N/A USAR Affected Pages: Section 8.3 Systems Reactor Building Lighting (LAR), Control Building Lighting (LAC), Reactor Building Drains (DER)
Title of Change: Revision to the NMP2 USAR Tables 8.3-16, Primary Containment Electrical Penetrations and Penetration Conductors Which Are Not Required During Reactor Operation and Are Protected by De-energization, and 8.3-17, Overcurrent Protective Devices for Nonclass 1E Lighting
- Fixtures on Class 1E Emergency System Description of Change:
This change revised USAR Table 8.3-16 to reflect as-installed.
plant conditions for those circuits associated with electrical penetrations and electrical penetration conductors. which are required to be de-energized during reactor operation. In addition, USAR Table 8.3-17 was revised to remove extraneous information.
Safety Evaluation Summary:
This change revised Table 8.3-16 to include a lighting circuit which is required to be de-energized for electrical penetration and penetration protection during reactor operation, and removed those circuits which are no longer required to be de-energized due to plant configuration changes. Additionally, Table 8.3-17 was revised to remove superfluous information, such as breaker manufacturer and breaker size, which is not required for a tabulation of those lighting circuits which isolate safety- and nonsafety-related circuitry.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 71 of 93 Safety Evaluation No.: 92-089 Implementation Document No.: N/A Affected Pages: N/A
'SAR System: 115-kV Offsite Power Sources Title of Change: Temporary Change to Scriba Substation to'Jumper Either Breaker R50, R60, R115 or R225 Description of Change:
This evaluation addressed the effect of installing a temporary change at the Scriba Substation to bypass either breaker R50, R115, R60 or R225 if any one of them should fail. This bypass allows the offsite 115-kV source to be reenergized if one of the breakers should fail. One breaker for each source may be bypassed, but not both. If breaker R50 or R60 is bypassed, then the transfer trip signal will be wired to R115 or R225, respectively.
Safety Evaluation Summary:
The existing Unit 2 protection schemes will not be changed. The only difference is which breaker in the Scriba Substation will be tripped by the transfer trip scheme from Unit 2. The reliability and protection of the two offsite sources for Unit 2, lines 5 and 6 (sources A and B), will not be affected.
Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 72 of 93 Safety Evaluation No.: 93-001 Implementation Document No.: Temporary Mod.93-005 USAR Affected Pages! 10.4-20, Figures 10.4-8 Sh. 8, 10.4-7d System! Circulating Water System (CWS)
Title of Change: Bypass of Failed Cooling Tower Basin RTD Description of Change:
This evaluation addressed the temporary removal of a failed cooling tower temperature detector. Signal wires from the defective cooling tower basin water temperature detector were lifted to prevent. potential spurious automatic Mode IV operation (warm water bypass) of the cooling tower.
Safety Evaluation Summary:
The two-out-of-four low basin water temperature logic which automatically initiates the Mode IV operation of the cooling tower to a two-out-of-three logic was found acceptable. The change prevented a faulty instrument, from spuriously initiating the automatic function. Operations personnel still maintained adequate monitoring and control capabilities of the instruments, temperatures, and system.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page 73 of 93 Safety Evaluation No.: 93-006, Rev. 1 Implementation Document No;: Specification P304SA, Procedure NMP2-IST-001 Rev. 3, 2PPD-GL-89-10 Rev. 1 USAR Affected Pages: Section 3.9A System: DBA Hydrogen Recombiner System Title of Change: Update USAR.Table 3.9A-12 "Active Valves (BOP)"
Description of Change:
This change revised USAR Table 3.9A-12, list of active valves,.to agree with descriptions in the USAR; with NMP2-IST-001 Rev. 3, "Pump and Valve First Ten-Year In-service Testing Program Plan"g and 2PPD-GL-89-10 Rev. 1, "Motor-Operated Valve Program Plan Description for-Nine Mile Point Nuclear Station Unit g2."
Safety Evaluation Summary:
Table 3.9A-12 was revised to change the valve type and manufacturer for 2HCS*MOVlA, 2HCS*MOV1B, '2HCS*MOV3A, 2HCS*MOV3Bg 2HCS*MOV4A, 2HCS*MOV4B, 2HCS*MOV6A, and 2HCS*MOV6B from Velan globe to Westinghouse gate. Also, the valve operator model number for 2HCS*MOV1A, 2HCS*MOV1B, 2HCS*MOV3A and 2HCS*MOV3B was changed from SMB-000-5 to SMB-00-10. The table is now consistent with as-built plant conditions and design basis documents (design specification NMP2-P304SA).
This change has no impact on the safe operation or shutdown of the plant. No physical hardware changes are required; therefore, the function and operability of the valves is,not affected.
The change from globe to gate valves was previously reported in submittal letter dated April 28, 1989, under Safety Evaluation 88U-008.
Based on the evaluation performed, itsafety is concluded that question.
these changes do not involve an unreviewed
Safety Evaluation Summary Report Page 74 of 93 Safety Evaluation No.: 93-009 Implementation Document No.: Procedures GAP-POL-01 Rev. 01, NEP-POL-300 Rev. 01 USAR Affected Pages: Sections 13.1, 13.2 System! N/A Title of Change: Restructuring of Nuclear Support Organization Functions in Accordance with Revised Procedures GAP-POL-01 and NEP-POL-300 Description of Change:
Changes have been made to the corporate level management and t;echnical support structure of NMPC's Nuclear Division including:
reorganizing the Licensing Branch and Information Management Branch of the Nuclear Support Organization back*under the Nuclear Engineering Organization; reorganizing the Training Branch and Emergency Preparedness Branch of the Nuclear Support Organization back under the Nuclear Generation Organization; reorganizing the Procurement Branch of Nuclear Support under the Nuclear Generation Organization; dissolving the Nuclear Support Organization and eliminating the position of Vice President Nuclear Support.
Safety Evaluation Summary:
The new organizational structure provides for the integrated management of activities that. support the operation and maintenance of Nine Mile Point Unit 1 and Unit 2. The Vice President Nuclear Generation will have overall responsibility for the support functions of Training, Emergency Preparedness, and Procurement, in addition to his other responsibilities. The Vice President Nuclear Engineering will have overall responsibility for the support functions of Licensing and Information Management, in addition to his other responsibilities. These changes provide clear corporate management control/direction of onsite and offsite support functions. These changes allow for dissolving the- Nuclear Support Organization and eliminating the position of Vice President Nuclear Support.
Based on the analysis performed, the new organizational structure for the support functions of Licensing, Information Management, Training, Emergency Preparedness, and Procurement does not constitute an unreviewed safety question.
Safety Evaluation Summary Report Page 75 of 93 Safety Evaluation No.s93-014 Implementation Document No.s Process Control Program, GAP-POL-01, GAP-OPS-01 USAR Affected Pages: Sections 13.1, 13.5 Systems N/A Title of Changes Radwaste Operations Reorganization Description of Change:
Responsibilities for Radwaste Operations was transferred from the Manager Operations to the Manager Radiation Protection. The change provides single-point accountability for radioactive waste shipping and improves the efficiency of radwaste processing.
Safety Evaluation Summary:
The reorganization provides a closer work relationship between Radwaste Operators and Radiation Protection workers. This reorganization does not affect Technical Specifications, 'the safe operation, or the safe shutdown of the plant.
Based on the evaluation performed; it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 76 of 93 Safety Evaluation No.s93-015 Implementation Dooument No.: Temp. Mod.93-017 USAR Mfected Pages: Figure 10.1-6C System:
Title of Change: Removal of Reactor Water Feed Pump P1B, Pump Seal Vent Valves Description of Change:
This change relates to the reactor feed pump seal water system (FWP), which supplies seal water to the reactor feed pumps mechanical seals to minimize feedwater outleakage. Temporary Modification 93-017 documents and controls the temporary removal of valves 2FWP-V2000B and 2FWP-V2001B and the installation of threaded caps in their place until the valves are reinstalled.
Safety Evaluation Summary:
This review, which included the effects of the change on the system s operability, reliability, maintainability, structural integrity, and system interactions, has found that the implementation of this change will not cause any significant safety or operability issues. They can be loosened to vent any entrapped air and, due to their weight, have no effect on structural integrity.
Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.
this
Safety Evaluation Summary Report Page 77 of 93 Safety Evaluation No.: 93-016 Implementation Document No.: Procedure NIP-TQS-01 USAR Affected Pages: Section 13.1 System: N/A Title of Change! Changes to NIP-TQS-01 to Describe Nine Mile Point Unit 1 and Nine Mile Point Unit 2 Staff Positions Comparable to ANSI N18.1-1971 and ANSI/ANS 3.1-1978 Description of Change:
This change cross-references titles used for staff members at Nine Mile Point Unit 1 and Nine Mile Point Unit 2 to comparable positions as shown in ANSI/ANS 3.1-1978.
Safety Evaluation Summary:
This change clarifies the staff member titles and their qualifications as required by ANSI/ANS 3.1-1978 and as committed to in the Technical Specifications and USAR.
Based on the evaluation performed, does not involve an it unreviewed is concluded that this safety question.
change
Safety Evaluation Summary Report Page 78 of 93 Safety Evaluation No.: 93-021, Rev. 1 Implementation Document No.: Calculations HVC-072 and HVC-073 USAR Affected Pages: Section 9.4 System! Control Building Unit Coolers Title of Change: Technical Specification Interpretation f25, USAR and EQD Changes Description of Change:
USAR Table 9.4-1 was revised to reflect new design temperatures to allow unit coolers 2HVC*UC104, 105, 106 E 107 to be out of service without imposing an LCO. The equipment located in the zones affected by unavailability of the above unit coolers will either continue to function, or and will not interfere with if it other fails, safety it will fail safe functions. If any one of the unit coolers (2HVC*UC104, 105, 106, 107) is inoperable, the temperature of the affected areas must be monitored. If the temperature exceeds 1044F, an equipment operability review shall be required.
Inoperability of unit coolers 2HVC*UC103A/B requires LCO 3.7.3 (seven days). In any operational condition, the associated chiller should be taken .out of service if unit cooler 2HVC*UC103A or B is out of .service.
The proposed Technical Specification Interpretation (TSI) f25, Rev. 11, can be used for the plant operation.
Safety Evaluation Summary:
Technical Specification Interpretation f25 allowed control building unit coolers 2HVC*UC103A/B and 2HVC*UC104, 105, 106 and 107 to be out of service based on the fact that the areas cooled by these unit coolers were analyzed for loss of cooling and that the consequences will be within the design limit. Further review indicates that 2HVC*103A/B are required to be operable. Unit coolers 2HVC*UC104, 105, 106 and 107 can be taken out of service without entering an LCO. The following provides the basis for the above changes:
In areas cooled by 2HVC*UC104, 106 and 107, that, although temperatures may exceed the it was determined maximum design temperature when the unit coolers are out of service and LOCA
Safety Evaluation Summary Report Page 79 of 93 Safety Evaluation No.: 93-021, Rev. 1 (cont'd.)
Safety Evaluation Summary: (cont'd.)
occurs, equipment will either withstand the higher temperatures, or fail safe, or will have accomplished their safety function prior to exceeding the temperature for which they are qualified.
In 230'lectrical tunnel (Div. II), cooled by 2HVC*UC105, determined that the maximum design temperature of 104'F will it was not be exceeded if the unit cooler is out of service and LOCA occurs.
During normal plant operation with unit coolers operating, the area temperatures are lower than the above-defined LOCA temperatures. Therefore, equipment operability is not impacted.
However, if during normal plant operation any safety-related unit cooler (2HVC*UC104, 105, 106, 107) should be inoperable, and the area temperature exceeds 104'F, an engineering evaluation will be required to ensure equipment operability.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page 80 of 93 Safety Evaluation No.: 93-024 Implementation Document No.: DER 2-92-4027 USAR Affected Pages': Section 9A.3 System: Fire Protection Water (FPW)
Title of Change: 24 Vdc Starting Batteries Operability Criteria Description of Change:
Batteries 2FPW-BAT1A and 1B are used to start diesel-driven fire pump 2FPWQ-P1. This change revised the surveillance requirement of verifying. every 92 days that the difference in specific cells does not exceed.0.015, gravity of the electrolyte between acceptance to a new value of 0.040. The new criteria of 0.040 is based on the battery manufacturer's recommendation to ensure sufficient charge of the cells to start the fire pump. The pump manufacturer, who established the original 0.015 criteria, concurred with the change to 0.040.
Safety Evaluation Summary:
This change establishes a more realistic criterion of 0.040 for the specific gravity difference for the diesel-driven fire pump starting batteries. The new criterion ensures a sufficient degree of charge and the operability of the batteries to provide proper starting of the diesel fire pump. The fire protection system continues to function as designed, ensuring the availability of equipment important to safety in the event of a fire.
Based on the evaluation performed, it safety is concluded that question.
this change does not involve an unreviewed
Safety Evaluation Summary Report Page 81 of 93 Safety Evaluation No.: 93-025, Rev. 1 Implementation Document No.: Procedure N2-OP-53A USAR Affected Pages: N/A System:
Title of Change: Resolution of DER 2-93-0032 Description of Changes Operating procedure N2-OP-53A required that when an operating cooling unit in the control room or relay room is put into Pull-To-Lock (PTL) mode to test operability, of the redundant unit, the operating unit is declared inoperable. This is not required as long as cooling can be restored in 10 minutes.
Safety Evaluation Summary:
For situations described in DER 2-93-0032, the time to reach the design temperature of 90'F is calculated to be about 13 to 15 minutes. If cooling is restored within 10 minutes, then the temperature within either the control room or relay room will be within the design limit, and component operability is not impacted.
Therefore, Operating Procedure N2-OP-53A was revised such that the HVC*ACUs may be placed in PTL for 10 minutes without declaring the operating unit inoperable.
Based on the, evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 82 of 93 Safety Evaluation No.: 93-026 Implementation Document No.: Generic Letter 89-13 USAR Affected Pages! N/A Systems HVC Title of Change: Performance Testing of Unit Coolers Description of Changers Generic Letter 89-13 requires that heat exchangers cooled by service water be regularly tested to evaluate their design capacity. This safety evaluation addresses testing of the unit coolers using service water.
Since the unit coolers do not have flow instrumentation permanently installed, a temporary test loop will be connected using fire hose. The test loop will be installed in place of spool pieces installed for flushing and testing unit coolers.
Test results will be analyzed to provide repositioning of service water flow control valves where necessary.'afety Evaluation Summary:
This test program adds a test loop to a unit cooler. The addition of the test loop is only for the duration of each test and no other changes (pressure, piping arrangements) are required. Localized flooding from rupture of the test loop is bounded by previous flooding analysis. Therefore, the test program does not impair the safe operation of the plant.
Based on the evaluation performed, it is concluded that this safety question.
change does not involve an unreviewed
Safety Evaluation Summary Report Page 83 of 93 Safety Evaluation No.: 93-031 Implementation Document No.: Procedures NEP-POL-300, NIP-IRG-01, NIP-ECA-04 USAR Affected Pages: Section 13.1 System: N/A Title of Change: Nuclear Licensing Organizational Structure and Responsibilities Revised Procedures NEP-POL-300, NIP-IRG-01 and NIP-ECA-04 Description of Change:
The organizational structure of the Nuclear Licensing Organization has changed such that, the Manager Licensing reports directly to the Executive Vice President Nuclear. Prior to this change, the Manager Licensing reported directly to the Vice President Nuclear Engineering. In addition, the Manager Licensing has assumed the responsibilities for interfacing with INPO, and implementing the Quality First Program. These responsibilities were transferred from the Manager Executive Staff. The Manager Executive Staff position has been eliminated.
Safety Evaluation Summary:
The changes made to the organizational structure of the Nuclear Engineering and Nuclear Licensing Organizations continue to provide for the integrated management of activities that support the operation and maintenance of Nine Mile Point Unit 1 and Unit
- 2. These changes also continue to provide clear management control and effective lines of authority and communications between the organizational units involved in the management, operation, and technical support of the operation of Nine Mile Point Unit 1 and Unit 2.
Based on this evaluation, the organizational structure of the Nuclear Engineering and Nuclear Licensing Organizations continues to satisfy the acceptance criteria of SRP 13.1.1, and does not constitute an unreviewed safety question.
Safety Evaluation Summary Report Page 84 of 93 Safety Evaluation No.: 93-033, Rev. 1 Implementation Document No.: Procedure N2-OP-94 USAR Affected Pages: Section 6.2 System: Traversing In-Core Probe (TIP)
System Title of Change: Treatment of TIP Leakage in LOCA Analysis Description of Change:
This change revised Note 19 of USAR Table 6.2-56 as follows:
1~ The statement that the TIP. leakage path considered in the Chapter 15.6.5 LOCA analysis accounts for a break in a TIP line "occurring simultaneously with or because of a design basis LOCA>> was deleted.
2 0 The statement that only one of five TIP ball valves is open at a time to conduct LPRM calibrations was revised to indicate that a maximum of five valves may be opened at any one time.
Safety Evaluation Summary:
- 1. Deleting the statement that the TIP line break occurs with
. or because of a design basis LOCA removed an editorial change made in USAR Rev. 0 by LDCN U-238. LDCN U-983 made additional changes to Note 19 in the same revision, with substantial technical explanation. LDCN U-983 indicated that the same statement revised by LDCN U-238 should be deleted. Changes to the same statement from both LDCNs were inadvertently incorporated into Note 19. The deletion of the statement resolves this prior inconsistency. Leakage via a broken TIP guide tube is still considered in the LOCA analysis.
2 ~ NEDC-22253, >>BWROG Evaluation of Containment Isolation Concerns," October 1982, and. NMP2 compliance with the NEDC criteria, have been accepted by the NRC. This safety evaluation summarized three topics in particular from -the NEDC-22253 analysis that provide the basis for allowing all five TIP ball valves to be open at a time event probability, instrument line break considerations, and the LOCA and failure of five TIP guide tubes.
Safety Evaluation Summary Report Page 85 of 93 Safety Evaluation No.: 93-033, Rev. 1 (cont'd.)
Safety Evaluation Summary: (cont'd.)
These changes did not require plant modifications. Operating procedure N2-0P-94, "Traversing Incore Probe", changed step D.16 to allow a maximum of five ball valves to be opened at a time.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 86 of 93 Safety Evaluation No.!93-034 Implementation Document No.: Procedures GAP-POL-01, NEP-POL-300 USAR Affected Pages: Sections 9A.3, 13.1, 13.5, Appendix B Systems N/A Title of Changers Restructuring of Nuclear Generation and Nuclear Engineering Organizations per Revised Procedures GAP-POL-01 and NEP-POL-300 Description of Change:
The Nuclear Generation and Nuclear Engineering organizations are controlled administratively by procedures GAP-POL-01 and NEP-POL-300, respectively. These procedures establish functional positions and responsibilities necessary to ensure the safe and efficient operation, maintenance, modification, and testing of the Nine Mile Point Nuclear Station and are described in USAR Chapter 13. The restructuring expands the existing Site Services organization to include Nuclear Security, Technical Services (including Fire Protection, Central Maintenance, Environmental Protection, and Procedures), Procurement, and Construction Services. Relevant procedural changes, USAR changes, and NRC approved Plan and Program changes required to implement the new organizational structure have been made.
Safety Evaluati.on Summary:
The new organizational structure provides for the integrated management of common activities to support the operation and maintenance of the Nine Mile Point Nuclear Station. This organizational change altered the reporting structure of previous existing positions but. does not affect the performance of functions or responsibilities. The new reporting structure provides clear management control and effective lines of authority and communications between the organizational units involved in the management, operation, and technical support for the operation of the facility. These changes meet the acceptance criteria of Branch Technical Position CMEB 9.5.1, Standard Review Technical Specification 6.2.1. Based on Plan Chapter 13.1, and the evaluation performed, it is concluded that the restructured organization does not involve an unreviewed safety question.'
Safety Evaluation Summary Report 87 of 93 'age Safety Evaluation No.: 93-049 Implementation Document No.: Appendix.J Program Review USAR Affected Pages: Section 6.2 System: Residual Heat Removal Title cf Change: Venting of Shutdown Cooling, Containment Spray and .
Suppression Pool Cooling During the ILRT Description of Change:
This change corrects a discrepancy between the Appendix J Type A Test (ILRT) configuration and the USAR description. The shutdown cooling, containment spray and suppression pool cooling subsystems of the residual heat removal system have been added to the list of systems in the USAR which penetrate primary containment that may not be vented to the primary containment atmosphere during the ILRT. This change provides consistency between the Unit 2 Appendix J Program and the USAR while taking credit for an exception provided by 10CFR50 Appendix J.
Safety Evaluation Summary:
Appendix J provides for an exception to the venting requirements for systems that are required to maintain the plant in a safe condition during the ILRT, and for those systems that are normally filled with water and operate under postaccident conditions. This .change takes credit for the Appendix J exception and has .added shutdown cooling, containment spray and suppression pool cooling to the appropriate list in the USAR of systems which may not be vented during the ILRT., This change does not impact'the capability of these systems to achieve or maintain safe shutdown of the plant.
Based on the evaluation performed, change does not involve an it unreviewed is concluded that this safety question.
Safety Evaluation Summary Report Page 88 of 93 Safety Evaluation No.: 93-072 Implementation Document No.: LER 93-04 USAR Affected Pages! Section 6.2 System: Traversing In-core Probe (TIP)
Title of Change: Type B Testing of TIP Bellows Description of Change:
Appendix J of 10CFR50 requires a Type B test for the traversing in-core probe (TIP) penetrations Z31A-E due to the metal bellows arrangement. It was revealed during an independent review of the NMP2 Appendix.J program that these penetrations have not been Type B tested but have been included in the Type A test. LDCN 1458, dated November 29, 1984, added Note 34 to USAR Table 6.2-56, which stated that due to the metal bellows arrangement on TIP drywell penetration flanges, they will be included in Type A testing rather than Type B testing. This note was interpreted to apply to the testing of the metal bellows themselves. Therefore, due to this note, the TIP penetration bellows were not Type B tested but were included in the Type A test. However, no formal exemption request from Appendix J Type B test requirements was generated. It was incorrectly assumed when the operating license was issued that the USAR assumption of Type A testing was acceptable. The USAR has been revised to correctly identify Type B testing for the TIP penetration bellows.
Safety Evaluation Summary:
The Standard Review Plan (NUREG-0800) states that the primary containment leak testing program, as described in the USAR, will be acceptable if it meets the requirements of 10CFR50'ppendix J.
As stated in paragraph II.G.1 of Appendix J, containment penetrations, whose design incorporates piping penetrations fitted with expansion bellows and electrical penetrations fitted with flexible metal seal assemblies, are included in Type B testing. Conformance with the requirements of Appendix J constitutes an acceptable basis for satisfying the requirements of General Design Criteria 53 as it pertains to penetrations having resilient seals and expansion bellows. The TIP penetration bellows are being included in Type B testing as well as Type A testing to meet the requirements of 10CFR50 Appendix J.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 89 of 93 Safety Evaluation No.: 93-084 Implementation Document No.: NMP2-RG197-01 USAR Affected Pages! Section 7.5 System! Postaccident Monitoring and Display Instrumentation Title of Change: Revision of Unit 2 Final Safety Analysis Report (Updated) Subsection 7.5.2.1 and Associated Table 7.5-2 Description of Change:
USAR Table 7.5-2 was revised to contain only (1) a plant-specific listing of the variables for which postaccident monitoring instrumentation is to be provided at NMP2, and'2) the identification of variable type(s) and the designation of instrument category for each of these variables. Instrument-specific data such as component EPN numbers, instrument scale ranges, and instrument loop power supplies, have been deleted from the table. All deviations from instrument category designations, and all deviations 'from instrument design and qualification criteria recommended by RG 1.97 for each of the listed variables, continues to be explicitly identified and
)ustified in referenced notes.
The plant-specific list of RG 1.97 variables, and the identified type for each, are unchanged. With the exception of neutron flux variables, the instrument category specified for each RG 1.97 variable is unchanged.
Changes made to USAR Section 7.5.2.1 reflect the revised scope and content of revised Table 7.5-2.
Safety Evaluation Summary:
The level of detail provided in the revised table is still sufficient for reviewers to assess NMP2 conformance with the recommendations of RG 1.97.
Changes to the designated category for neutron flux instrumentation are consistent with information previously identified to the NRC by NMPC in Letter No. NMP2L-1394, dated June 18, 1993.
Safety Evaluation Summary Report Page 90 of 93 Safety Evaluation No.!93-084 (cont'd.)
Safety Evaluation Summary: (cont'd.)
The instrument-specific data which was deleted from the USAR has been relocated to a new NMPC document, NMP2-RG197-01, "Important Design Features of Regulatory Guide 1.97 Instruments for Nine Mile Point Unit 2." The changes to USAR Section 7.5.2.1 and Table 7.5-2 were implemented upon approval of NMP2-RG197-01.
These changes did not physically add any new instrumentation, nor were any existing plant structures, systems or components deleted or in any way physically modified. Implementation of the changes did not require changes to Technical Specifications. Also, implementation of the changes did not require changes to any operating instructions or to any maintenance or calibration instructions currently specified in any existing plant procedures.
Implementation of these changes has no adverse impact on the safe operation or shutdown of the plant.
Based on the evaluation performed, it is concluded that these safety question and are in changes do not involve an unreviewed full compliance with NRC standards.
Safety Evaluation Summary Report Page 91 of 93 Safety Evaluation No.: 93-086 Implementation Document No.: N/A USAR Affected Pages: Section 9A.3 System: N/A Title of Change: 1993 Fire Hazards Analysis Update Description of Changes This change revised fire loading table information to show updated calculation results and add transient combustible allowance.
Safety Evaluation Summary:
This change revised information in the fire hazards analysis summary tables due to updated calculation results and the addition of an allowable transient combustible loading factor to account for procedurally-controlled transient combustibles. The fire hazard analysis, which is performed in accordance with 10CFR50 Appendix R, Section II.B, and BTP CMEB 9.5-1, Position C.1.b, continues to verify that the fire hazards associated with Unit 2 have been appropriately considered. The revised analysis does not impact the ability to safely shut down the plant in the event of a fire, and no Technical Specifications are impacted.
Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.
this
Safety Evaluation Summary Report Page 92 of 93 USAR TEXT, TABLE AND FIGURE CHANGES (BASED ON PREVIOUSLY REPORTED SAFETY EVALUATIONS)
A number of text and figure revisions were made to the USAR to include additional changes that are based on previously reported safety evaluations. These changes are identified below.
Safety Evaluation No.: 87-080 Previously Reported: 10/26/88 The power supply to the HPCS diesel air compressor motor (2EGA-C3) was changed from Class 1E to non-1E. The following additional USAR figure has been updated accordingly:
Figure: 9.5-40a Safety Evaluation No.: 88U-077, Rev. 1 Previously Reported: 6/27/89 Standby diesel generator room emergency-duty outdoor air ventilation system air flow switches for applicable exhaust fans were previously relocated to the intake side of the fans after determining the location at the discharge produced inaccurate monitoring capability. The following additional USAR section and table have been updated accordingly:
Section: 7.3 Table: 7.3-16 Sh 1 Safety Evaluation No.: 89-044, Rev. 1 Previously Reported: 10/29/92 A set of stairs with'landing replaced the ladder for the el.
320~-3 1/4" platform used for accessing the screenwell building HVAC equipment room. The following additional USAR figure has been updated accordingly:
Figure: 12.3-69 Sh 4
Safety Evaluation Summary Report Page 93 of 93 USAR TEXT, TABLE AND FIGURE CHANGES (BASED ON PREVIOUSLY REPORTED SAFETY EVALUATIONS)
(Cont'd.)
Safety Evaluation No.: 8$ -075, Rev. 5 (currently Rev. 8)
Previously Reported: 10/30/91 Additional modifications to the Gaitronics communications system have been completed in accordance with Modification PN2Y87MX038, as described in Safety Evaluation 89-075. The following USAR section and figures have been updated accordingly:
Sections: 9.5 Figures: 9 5 5 Sh 1 & 2 9 5 6g 9 5 7 9 5 8 Sh 2g 9 5 9
~ ~ ~ ~ ~ Sh 2 9.5-10 Sh 2 9.5-11; 9.5-14; 9.5-15( 9.5-17 9.5-18 9.5-20 Sh 1/ 9 '-21) 9 '-24 9 '-29 9.5-33 9.5-34; 9.5-36 Safety Evaluation No.: 90-096, Rev. 1 (currently Rev. 2)
Previously Reported: 10/30/91 Additional modifications to the nuisance annunciator windows have been completed in accordance with Modification PN2Y86MX085. The following additional USAR figures have been updated accordingly:
Figures: 10.1-9a; 10.1-9b; 10.1-9c
e