ML20247B032
| ML20247B032 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/31/1988 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 1699, NUDOCS 8909120320 | |
| Download: ML20247B032 (95) | |
Text
g TOLEDO
%ms EDISON A Centenor Energy Company DONALD C. SHELTON Va Prment-Nor I*'****"
Docket Number 50-346 License Number NPF-3 Serial Number 1699 September 1, 1989 United States Nuclear Regulatory Commission Document Control' Desk Vashington, D. C. 20555
Subject:
1988 Annual 10CFR50.59 Report of Facility Changes, Tests and Experiments Gentlemen:
The Toledo Edison Company hereby submits, pursuant to 10CFR50.59(b)(2), the 1988 Annual 10CFR50.59 Report of facility changes, tests and experiments for
' Davis-Besse Nuclear Power. Station, Unit Number 1.
Those changes, tests and experiments identified via the safety review process during the reporting period of January 23, 1988 through January 22, 1989 are enclosed.' Attachment 1 provides an executive summary of those changes, tests and experiments contained in the enclosure.
Very truly yours, l
JCS/dlm Enclosure ces P. H. Byron, DB-1 NRC Senior Resident Inspector A. B. Davis, Regional Administrator, NRC Region III (2 copies)
T. V. Vambach, DB-1 NRC Senior Project Manager 8909120320 881231 PDR ADOCK 05000346 I t R PDC THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO OHIO 43652
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- J Licznza Nu1bar NPF-3 i-?!. -l L' -Serial Number 1699' Attachment!11 i
- Page 1 ATTACHMENT le 10CFR50.59
SUMMARY
SHEET' NUMBER < TITLE-i
. DCR 88-0030 Deletion of Fire Protection and Potable Water. Lines to Service Building Number 1
'DCR 88-0077 Valve Packing Material / Configuration Modifications
- DOR 88-0129
~
Revise Design Drawings to Reflect the Fire Protection Line to the New Flammable Material Storage Facility
.. . DCR 88-0158. Battery Load: Profile DCR 88-0226' r Replacement of Flow Glass with Isolation;
- Valve in~the Condensate Demineralized System-
' DCR 88-0255 Containment Spray. System USAR Drawing Change-FCR 79-0208, Supp. 12 New Radvaste Demineralized Unit
' FCR 84-0002 Reactor Vessel Head to Hot Leg Vent FCR 84-0044A' Room 428 Fire Dampers
- FCR 84-0049' Safety Grade Incore Thermocouple FCR 84-0051, Rev. 1 Lov Level Radioactive Vaste' Storage Facility
- FCR 84-0067, Rev. B Refuel Reactor FCR 84-0068, Supp. 1 Moisture Separator / Reheater Belly Drain Modification FCR 84-0094 Fire Barrier Between Room 123 and Room 240 FCR 84-0104, Supp. 1 Modify Nitrogen System for Electrical Penetrations FCR 84-0116, Supp. 4 Ex-Core Neutron Flux Monitoring System FCR 84-0190, Rev. B,~Supp. 6 Align AF-3872 Open
, FCR 84-0209 Fire Barrier at Vest Vall of Room 516 FCR 85-0065, Rev. 0, Supp. 5 Align AF-3870 Open FCR 85-0091 Control Room Dust Control
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-Dock $t Nurbar 50-346 ,j 1
-Licensa Numbar NPF-3 ..
" I
. Serial Number 1699' Attachment 1 Page 2 {
'FCR 85-0096, Rev. B Redundant Steam Generator Level and Pressure '
Indication FCR 85-0109 Install New Cabinet in Control Room
-FCR 85-0124 Upgrade Steam Vent Mixing Condenser
'FCR 85-0154, Rev. A, Supp. 2 Remove SFRCS Signals from AF-599 and AF-608 FCR 85-0157, Rev. A, Supp. 1 Lower SFRCS SG Lov and High. Lev'el Trip Setpoints FCR 85-0198, Rev. B CVRT Permanent Filter Installation FCR 85-0207, Rev. 1 CRDM Ventilation System FCR 85-0221, Supp. 3- Modifications to Service Water Lines for the Control Room Emergency Ventilation System I (CREVS) Condenser FCR 86-0036 RCS Extended Range Pressure Indicator
) FCR 86-0123 Reactor Power Auctioneer to the ICS FCR 86-0192, Supp. 2 Repower the Startup Feedvater Pump FCR 86-0220A Upgrade Fire Dampers
! FCR 86-0226A Upgrade Fire. Dampers FCR 86-0291 Modify High Pressure Injection System Flov l
Test Line FCR 86-0318 Relocate Hot Leg Level Monitoring System Reference Leg FCR 86-0330, Rev. B Auxiliary Feedvater Level Control FCR 86-0334, Supp. 2 Modify Station Air Compressor Pressure Safety Valve PSV 2119 FCR 86-0425, Supp. 8, Rev. 2 Motor Driven Feedvater Pump (MDFP) Phase II FCR 86-0432, Supp. 7 Enhanced Feed and Bleed Capability l
FCR 86-0432, Supp. 8 Enhanced Feed and Bleed Capability (Close MU 6421 and MU 6423B)
FCR 87-0063, Rev. 1 Remove SFAS Actuation Signals FCR 87-0064 Relocation of AFPT Isolation Valve Controls FCR 87-0065 AFV Isolation Valves SFRCS Block Switches
m, L LA , Dock t Numbsr 50-346' LLicenza Nulb2r NPF-3
. Serial' Number 1699
- Attachment 1
- Page 3.
, FCR 87-0067- Control Room Center Console H FCR 87-0069 Auxiliary Feedvater Flow Indication FCR 87-0070 Control' Room Center Console FCR 87-0071,^Supp. 1 Relabel Manually Operated System Bypass Status Indicating Light IL/HS 4808 FCR 87-0092,_Rev. A and Modification of AFV/SFRCS Manual FCR.87-0130; Initiation Switches FCR 87-0092,- Rev. A, Supp. 1- Modification of AFV/SFRCS Manual Initiation Switches (Provide Redundant Solenoid Valves for the Main Feedvater Control Valves)
FCR 87-0109 RPS High Pressure Setpoints and ARTS Threshold Power r ~ MOD 87-1011 Add Local Flow Indication to the AFV Recirculation Flow Lines MOD'87-1031- ' Rod Stop Circuitry MOD 87-1062 Upgrade Station and Instrument Air System
' MOD 87-1092 ICS/NNI Auto-Select Switch-MOD 87-1093, Supp. 1 Control Rod Drive System Improvements MOD 87-1107, Supp. 5 Decay Heat Removal Task Force and Systems Review and Test Program SFRCS Changes MOD 87-1124 Control Rod Drive Cooling Vater System Modifications.
MOD 87-1142 Generator Trip for Stator Ground Fault MOD 87-1148 HPI and DH/LPI Flow Indicators MOD 87-1168 Decay Heat Removal System Valve Replacement MOD 87-1195 Modify ECCS Room Sump Pump Piping and Conduit MOD 87-1260' Add Condensate Removal Capability 1
MOD 87-1261 RPS Conduit Not Seismically Supported MOD 87-1290 Remove the Internals of SV-329 MOD 87-1305 Modify Extraction Steam Control Valves
' Dock t Nuxb:r 50-346 Lic;nsa Numb:r NPF-3 Serial Number 1699 Attachment 1-Page 4 MOD 87-1317 345KV Generator Breaker Flashover Protection
' MOD 87-1318 Generator Inadvertent Energization MOD 88-0049 -Orient Valve MS 5889A to a Vertical Position MOD 88-0195 SFAS Test Circuitry Wiring MOD 88-0234 Overpressure Protection for the Containment Air Coolers (CAC)
MOD 88-0236 Whip Restraints on the 10" Pressurizer Surge Piping MOD 88-0265 Turbine Bypass Valve (TBV) Line Condensate Control l
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SAFETY EVALUATION
SUMMARY
, FOR.
TITLE:
-Deletion of Fire Protection and Potable Water Lines tc Service Building Number 1 CHANGE:n
' Revise USAR Figure-9.5-1 (P&ID M-016A) to show the Fire Protection water supply to Service Building Number 1 isolated and capped.
REASON FOR CHANGE:
Service Building Number'1 was dismantled and removed.
SAFETY EVALUATION
SUMMARY
The deletion of the fire protection piping to Service Building #1 vill have no impact.on the safety of the plant. This building served no safety related
-function, and this Temporary Mechanical Modification (TMM) is used to cap the piping which previously served Service Building #1 and does not otherwise affect the yard fire loop.
The domestic water system does not perform any functions which are important
-to safety.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type.than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
LTherefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
TITLE:
Valve Packing Material / Configuration Modifications CHANGE:
Revise USAR Sections 5.2.4.7, 5.5.12,-and 9.3.4.3.4 to remove the statement that leakage detection from RCS valve stems is used to determine a source of RCS inventory reduction and that reactor coolant pressure boundary (RCPB) valves are provided with double packing glands and lantern rings.
l REASON FOR CHANGE:
The Davis-Besse valve packing improvement program has deleted the use of double packing glands and lantern rings on all RCPB valves. In addition, the leakoff-connections between the double packing glands at the lantern ring are
-all plugged, have never been piped'to a sump and have never been utilized to identify RCS inventory reduction.
SAFETY EVALUATION
SUMMARY
This DCR is documenting changes only to valve packing glands. No changes are being made to any valve actuation' devices (i.e., motor, air, mechanical, etc).
Given this, these packing modifications have no impact on the ability of the l- valve to perform its intended safety related function during any postulated Design Basis Accident.
l l As summarized above, the proposed action will not increase the probability or I consequence-of an accident or malfunction previously evaluated in the USAR.
l The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR 9 DCR 88-0129 (SE 88-0535)
' TITLE:
Revise Design Drawings'to Reflect the Fire Protection Line to the New Flammable Material Storage Facility CHANGE:
Revise USAR Figure 9.5-1 (P&ID M-016A) to show the connection to the Fire Water Line for the Flammable Material Storage Facility.
.. REASON FOR CHANGE:
To provide fire' protection to the new Flammable Material Storage Facility.
SAFETY EVALUATION
SUMMARY
This change vill'not adversely impact the capacity of the fire protection system because:
1.- Simultaneous fires in different buildings (separate fire areas) are not routinely postulated in the design of proprietary fire protection systems. Should multiple fires occur (although this is not postulated),
~t he isolation, valves could be used to prohibit the flow of water to the-least essential fire protection system that was operating. This would increase the amount of water available to the most critical systems. An example of this would be simultaneous fires in the cable spread room and.
the training center. If a lov vater pressure condition developed, the training' center system would be shut-off in order to provide adequate water pressure to the more vital cable spread room system. It should be noted that neither NRC requirements nor insurance guidelines require simultaneous fires in different buildings.
- 2. The' sprinkler capacity is estimated at 1000 gpm when activated. This is well within the capacity of the 2500 gpm Fire Protection Pumps.
The new connectf.ru is located outside of the protected area. This portion of the Fire Protection System is considered "Non-0" because it is not required to support the safe shutdown equipment at the plant.
As summarized above, the proposed action vill not increase the probability or
--consequence of.an accident or malfunction previously evaluated in the USAR.
I The proposed action vill not create the possibility for an accident or l: malfunction of a different type than any evaluated previously in the USAR, and does not' reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR
.DCR 88-0158 and UCN 88-0031 (SE 88-0192)
- TITLE
Battery Load Profile CHANGE:-
f
-Revise USAR Section 8.3.2.1.2 to incorporate the new load profile for the station battery..
- REASON FOR CHANGE:
The station battery load profile (design duty cycle) has been revised by-calculation C-EE-002-005 Rev. O in order to incorporate fifth refueling outage modifications and to more closely define'the scenario to which the batteries are sized. This includes a more rigorous accounting for breaker operation during the design duty cycle.
SAFETY EVALUATION
SUMMARY
Incorporation of.the new load profile does not. adversely affect plant safety.
Calculation C-EE-002-005, Rev. O vhich generated the new profile showed that the present batteries are properly. sized to meet the anticipated battery.
loading.. This' battery capacity calculation included the appropriate correction factors fc -temperature and aging.per IEEE 485.
Th'e previous load profile, per calculation C-EE-002-004 Rev. 1, contains amperage values that analysis has shown to be overly conservative. The analyzed design scenario is a Loss Of. Coolant Accident.(LOCA) followed by a
. Loss of!Offsite Power (LOOP) with the single failure of one Emergency Diesel Generator.
A more accurate load list for the station batteries does not detract from
' plant safety and vill allow an accurate reference point for future analyses.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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l SAFETY EVALUATION
SUMMARY
TITLE:
Replacement of Flow Glass with Isolation Valve in the Condensate Demineralized System CHANGE:
Revise USAR Figure 10.4-8A (F&ID H-008B) to replace FGCD1 vith isolation valve CD 99.
REASON FOR CHANGE:
The function of the flow glass, FGCDl, was to allow operators to observe flow to the Condensate Demineralized Precoat Tank during polisher precoating operations. Because of leaks in the system, FGCD1 was replaced with an isolation valve.
SAFETY EVALUATION
SUMMARY
This change does not impact a safety function and has no affect on any safety equipment.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than ary evaluated previously in the US/A. and m
does not reduce any margin of safety as defined in the Technica.'
Specifications.
Therefore, an unreviewed safety question does not exist.
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SUMMARY
TITLE:
Containment Spray System USAR Drawing Change CHANGE:
Revise'USAR Figure 6.3-3 (P&ID M-034) to show valves CS 1591'and CS 1592 as normally closed.
REASON FOR CHANGE:
Local containment' spray pump suction pressure indicators PI1591 and PI1592 are not capable of withstanding an overpressurization if valves DH 2733 or DH 2734 open while the. plant is in transition between normal operation and normal cooldown through the decay heat system. In order to prevent this occurrence, PI1591 and PI1592 are valved out utilizing containment spray pump suction pressure source valves CS 1591 and CS 1592. As a. result, valves CS1591 and CS 1592 are normally closed and are open only during testing of the containment spray pumps.
SAFETY EVALUATION
SUMMARY
The Containment Spray-System is used during post-accident conditions and serves no function during normal plant operation.
Pressure indicators PI1591 and PI1592 are not used while the system is operating and provide local indication of the pumps suction during testing.
The proposed change to isolate the indicators by closing valves CS 1591 and CS 1592.during non-testing periods vill reflect the'as-built condition of the plant and prevent possible overpressurization of the indicators if.the valves were left normally open.
The proposed change ensures that the pressure indicators will be available to perform their normal function. No changes to plant procedures resulted from this change.
As summarized above, the proposed action vill not increase the probability or
. consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or l- malfunction of a different type than any evaluated previously in the USAR, and l- does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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y SAFETY EVALUATION
SUMMARY
, FOR FCR'79-0208 SUP.'12 TITLE:-
Nev Radvaste' Demineralized Unit
'n . CHANGE:
' Prepare. space for'and provide services to a new radvaste deminexalizer unit to be located in'the.' Auxiliary Building Fuel Handling Area, Room 30U.
AEASON FOR' CHANGE:
The present'radvaste demineralized is.to'be removed.and' replaced by-a unit to.
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be supplied by Duratek.
-SAFETY EVALUATION
SUMMARY
The proposed change vill not affect safe operation <of the plant.- The demineralized: skid is not required for safe shutdown of. the plant, nor is:it located near'any safety-related equipment.
Shield panels surrounding the' demineralized skid are seismically supported for personnel protection, to prevent demineralized damage, and to permit. future installation.of safety,related equipment in the area of the skid if required.
EAs summarized above, the proposed action vill not. increase the probability or
. consequence of an accident or malfunction previously evaluated in the USAR.
'.The proposed action-vill not. create-the possibility for an accident or malfunction of a different type than any evaluated previously in the U$4R, and does not reduce any margin of safety:as defined in the Technical-S,ecifications.
Taerefore, an unreviewed safety quertion does not exist.
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SUMMARY
FOR cm .
FCR 84-0002:
< TITLE
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Reactor Vessel Head to Hot Leg Vent.
CHANGE: ,f Install a 2%" Schedule 160 line from the reactor' vessel head to the steam-generator. upper plenum _to provide a continuous vent of non-condensible gases
, from'the' Reactor Vessel.
REASON FOR CHANGES-To meet'the requirements.of 10CFR50.44 to install a reactor vessel head vent system.-
SAFETY EVALUATION'
SUMMARY
The safety function of the reactor vessel (RV) head to hot leg (HL) vent line is to vent the steam or non-condensible gas bubbles in the reactor vessel head-to the hot Icg where steam could.be condensed or non-conder.sible gases'could be subsequently vented by tlue high point vent. The other safety function of:
the:11ne is to reduce the temperature difference between the RV head and the-hot ~1eg by cooling the metal in the head during natural circulation. This
- vill reduce the thermal stress in the head and speed up the natural circulation cooldown process ('20*F/hr) and allow the hot.2eg level measure:nent system (HLLMS) to track the RCS inventory under natural circulation or SBLOCA conditions when the RCPs are tripped.
Tnis modification to the reactor coolant system will improve-the system
. operation without any adverse impact on the safety function of the RCS and
'other related systems'or components.
-As-summerized above, the proposed action vill not increase the probability or.
- consequence of an accident or malfunction previously evaluated in the.USAR.
LThe proposed action vill not create the possibility for an accident or malfunction _of a different type than any evaluated previously in the USAR, and
'does not reduce any margin of safety as defined in the Technical Specifications.
, Therefore, an unreviewed safety question does not exist.
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- S SAFETY EVALUATION SUMHARY FOR FCR B4-0044A TITLE:
Room 428 Fire Dampers CilANGE:
Replace fire dampers FD-1053 and FD-1055 REASON FOR CHANGE:
Dampers'FD-1053 and FD-1055 vere not originally installed per manufacturer's and NFFA Code requirements.
SAFETY EVALUATION S!!MMARY:
Modification of the fire dampers in the Auxiliary Building Non-Radvaste Area Ventilation System (FD-1033, FD-1055) vill not adversely affect the safety of
- the plant based on an HVAC system analysis because the system does not perform a safety function. The fire dampers do perform a safety function by virtue of their presence in fire barriers protecting redundant trains of safe shutdown i- (App. R) equipment.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluatea previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR FCR 84-0049 TITLE:
Safety Grade Incore Thermocouple CHANGE:
Replace sixteen existing incore thermocouple that are used in the ICC detectir.g systehi with rafety grade units and install environmentally qualified and separated cabling for the detection system.
REASON FOR CHANGE:
NUREG-0737 (Item II.F.2) and Regulatory Guide 1.97, Rev. 3 mandate that reactor incore thermocouple shall be used as part of an Inadequate core Cooling (ICC) detection system and that the thermocouple measuring channels must be Class lE. This means that they must be environmentally qualiffed in accordance with Regulatory Guide 1.89 (NUREG-0588) and physically separated in accordance with Regulatory Guide 1.75.
SAFETY EVALUATION
SUMMARY
The proposed exchange of sixteen (16) presently used Incore Detector Assemblies with improved units that have enviror. mentally and seismically qualified thermocouple vill improve the capability of providing reactor core exit temperatures to the operator. The installation of new environmentally qualified thermocouple extension cables that are separated into two (2) safety !
channels and leave the containment via two (2) new separate Class lE qualified penetration assemblies improves the reliability of the core exit temperature measurement.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and l
.does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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.V L .. SAFETY EVALUATION
SUMMARY
FOR FCR 84-0051, REV. 1
. TITLES <
'Lov Level Radioactive Vaste Storage Facility-
' CHANGE:- *
. Construct.a facility for the onsite storage of low level radioactive vaste.
. REASON FOR CHANGE:
To provide a facility that is designed to' accept and store'in a retrievable manner an acicamulated'five (5) years of vastes generated from the operation of
~
the Davis-Besse Nuclear Power Station Unit 1. The facility design allows for
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' future expansion of the storage area. The vastes are to be stored in the facility for no more than five years or until they can be shipped to a permanent disposal site as indicated in NRC Generic Letter 81-38.
SAFETY EVALUATION
SUMMARY
No safety-related components are related to or affected'by the operation of the Low Level Radioactiv'; 'iaste Storage Facility. No testing or experimentation vill be performed that will in any way differ or be less conservative than any previously described in the USAR. The facility is intenced solely for the storage of low-level radioactive vaste; no processing or experimentation vill be performed.
Accident' analysis demonstrate that technical specification limits for normal
.' radioactive releases are not approached by even the worst' case accidents that-are postulated for the storage facility. The operation of this facility vould in no way reduce the margin of safety defined in the basis of any Davis-Besse technical specification.
The radiological consequences of worst case unexpected accidents within the facility have been analyzed and indicats that any potential dose represents a-very small fraction of the 10CFR100 cid*ite dose criteria.
As summarized above, the proposed actlon vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical j Specifications.
Therefore, an unreviewed safety question does not exist.
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1 SAFETY EVALUATION
SUMMARY
FOR
.FCR 84-0057, REV. B;(SE 87-0288, REV. 1)
TITLE:
Refuel Reactor CHANGE:
' Refuel the reactor'for Cycle 6.
REASON FOR' CHANGE:.
LThis FCR provides the loading of new fuel assemblies (FAs) and burnable poison
, rod assemblies (BPRAs), the. shuffling of FAs and control rod assemblies (CRAs) and the. replacement of eight " black" axial power' shaping rods (APSRs) with
-eight " gray" APSRs to facilitate nuclear power generation for Cycle 6.in accordance vith the limits and analysis presented in BAV-2038, April 1988, Davis-Besse Nuclear Power Station, Unit Number 1, Cycle 6-Reload Report.
SAFETY EVALUATION
SUMMARY
The reference cycle for the nuclear and thermal-hydraulic design'of Cycle 6 is Cycle 5. 'The Cycle 6 physics parameters are based on a 400 effective full power day (EFPD) Cycle 5 length including APSR vithdrawal and coastdown.
There have been no anomalies during Cycle 5 which would adversely affect fuel performance during Cycle 6 as designed. Consistent with previous cycles, cross-core shuffling of. fuel assemblies is minimized in this cycle. The Cycle 6 design is characterized by only 16 FAs being cross core shuffled so as to minimize any carryover effects from tilts encountered in previous. cycles.
The Cycle 6 loading includes 64 new FAs (Batch 8) at 3.13 v/o U-235 and the reinsertion of one (Batch 1) previously discharged FA. This loading, characterized as Batch 8, is comprised of the MK-B5 design which is the same-as the Batch 7 design currently in'use. Due to the design length of Cycle 6 (405 EFPD), additional reactivity is necessary. This increased reactivity will be controlled, in part, by 64 new BPRAs (of the same physical design as used in Cycle 5)' located in the fresh fut ;. The reactivity is also controlled by soluble boron and 53 full-length AF-In-Cd CRAs. These CRAs are the same ones_used in previous eveles. However, the rod group designations differ from Cycle 5 in order to increase the vorth of group 4 to facilitate control during physics testing and to decrease the vorth of group 7 to be compatible with the new gray APSR imbalance control capability.
All accidents analyzed in Chapter 15 of the Updated Safety Analysis Report
.have been reexamined, with respect to Cycle 6 parameters, to ensure that the thermal performance during the hypothetical transients has not been degraded.
The hot full power moderator and Doppler coefficients remain negative such
.that Cycle 6 is bounded for main steam line break or any other over-cooling
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have been evaluated using conservative radionuclides source terms that bound the cycle specific source term for the longer Cycle 6 and future reload $
cycles. The results of the dose evaluations show that the offsite l radiological doses for each accident are below the respective acceptance criteria values in the current NRC Standard Review Plan NUREG-0800, i
As summarized above, the proposed action vill not increase the probability ..t l c consequence of an accident or malfunction previously evaluated in the USAR.
l The proposed action vill not create the possibility for an accident or 4
! malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR FCR 84-0068 SUP. 1 (SE 87-0255)
- TITLE:
Moisture Separator / Reheater Belly Drain Modif>cnt r?s CHANGE:
Reconfigure the Moisture Seperator Reheater (h :h) belly drain piping by 1) relocating the drain control valves (RD 2146, 2148, 2150, and 2151), 2) replacing the downstream carbon steel piping with ntainless steel, and 3) installing isolation valves upstream and downstream of the control valves and at the drain lines discharge to the condensate header.
REASON FOR CHANGE:
- 1. To locate the drain control valves closer to their condenser penetration.
- 2. To minimize pipe erosion which has caused numerous repairs.
- 3. To provide isolation valves to facilitate control valve maintenance and line isolation as necessary.
SAFETY EVALUATION
SUMMARY
No nuclear safety-related components are being affected by this change. None of these piping changes v'11 affect the function or the effectiveness of the MSR. No changes are being made that could alter the present condenser penetration such as to alter the effectiveness of the condenser. All changes are vell upstream of the condenser.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
l Therefore, an unreviewed safety question does not exist. '
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SAFETY EVALUATION
SUMMARY
FOR-FCR 84-0094-TITLE:-
Fire Barrier Between Room 123,and Room 240
' CHANGES-Install one fire damper'(FD-1192).in a duct and seal two pipe penetrations in the vall between Room 123 and Room 240.
REASON FOR CHANGE:
To' upgrade the vall;between Fire Area A (Room 123) and Fire Area G (Room 240)
'to a-three-hour' fire rated barrier.
SAFETY EVALUATION
SUMMARY
Sealing of the two pipe penetrations vill not affect the operability of any system, since the pipes vill be externally sealed and do not need to be isolated. Auxiliary 3uilding ventilation (specifically the ventilation of Room-123)'will be affected by the addition of fire damper FD-1192; the l principle effect being the partial' loss of ventilation during the installation
'of the fire damper.
None of the equipment: in Room 123 is required to operate during an accident condition, and no.maloperation could be caused by this modification.
As summarized above,.the proposed action-vill not increase the probability or consequence'of an accident or malfunction previously evaluated.in the USAR.
The' proposed action vill not create the possibility for an-accident or .
malfunction of a different type than any evaluated previously in the USAR, apd.
does not reduce any maxgin of safety as defined in the Technical-
. Specifications.
Therefore, an unreviewed safety question does not exist.
SAFETY EVALUATION
SUMMARY
FOR FCR 84-0104, SUPP. 1 TITLE:
Modify NitroFra System for Electrical Penetrations CHANGE:
Relocate the electrical penetration nitrogen storage bottles and allow the nitrogen supply bottles to remain connected to the system to provide a constant supply of nitrogen until the bottle is depleted.
REASON FOR CHANGE:
The storage bottles are being located in an area that is accessible post-LOCA.
The permanent connection vill provide a mea.s cf automatically maintaining the nitrogen pressure in the electrical penetration.
SAFETT EVALUATION
SUMMARY
The modification vill have no effect on safety because the nittogen bottle mounted in the auxiliary building is mounted Seismic Class I and provided with missile protection. The system is also provided with adequate relief capacity so the penetrations can not be overpressurized. Control of the bottles vill be defined by SP 1104.76, Electrical Penetration Nitrogen Blanketing.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
1 L - ___
y SAFETY EVALUATION
SUMMARY
FOR 1 FCR 84-0116, SUPP. 4 (SE 88-0172)
TITLE:
Ex-Core Neutron Flux Monitoring-System CHANGE:
Install two new Ex-Core Neutron Flux Monitoring Systems that are in addition to the currently installed Nuclear Instrumentation.
REASON FOR CHANGE:
.To provide neutron flux measurements that vill indicate whether ple.nt safety functions are being accomplished and provide information required to mitigate the consequences of an accident.
The new Ex-Core Neutron Flux Monitoring System will be installed to fulfill the requirements of NUREG 0737, Supplement 1.
SAFETY EVALUATION
SUMMARY
The new Ex-Core Neutron Flux Monitoring System vill meet all the requirements that are specified in Regulatory Guide 1.97. This means that all the new system components vill be:
(a) Environmentally qualified per IEEE 323-1974 (b) Seismically qualified per IEEE 344-1975 (c) Independently separated per IEEE 384-1977 and that the system cables, splices, and connections vill be qualified per IEEE 383-1974 The installation of qualified Ex-Core Neutron Flux Monitoring Systems in addition to the presently installed Nuclear Instrumentation System (which will not be removed) vill increase the safety aspects of the neutron flux measurements. Namely, the operator vill now have available to him two (2) additional neutron flux measuring channels, located at Post Accident Monitoring Panels C-5798 and C-5799, that consist of fully qualified components. This vill improve the confidence level of the information provided to the operator so that he can determine whether plant safety l
functions are being accomplished.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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L i SAFETY EVALUATION
SUMMARY
FOR . lf FCR 84-0190, Rev. B, Supp. 6 (SE 88-0007 Rev. 1)-
TITLE:
Align AF-3872 Open CHANGE:
l L . Align Auxiliary. Feed Fump 1-2 Discharge Valve AF-3872 from normally closed to-normally open, lock'the handwheel and install a cover plate on the local control station to prevent. inadvertent closure of the valve.
REASON FOR CHANGE:-
Aligning AF-3872 normally open together with the modifications implemented by FCR 85-0154,-FCR 85-0065 and FCR 86-0330 vill improve the reliability of the Auxiliary Feedvater System.
SAFETY EVALUATION
SUMMARY
After implementation of FCR 85-0154, FCR 86-0330, and FCR 84-0190, with'the plant in normal operation, the following condition vill exist: AFF 1-2 Discharge Valve AF 388 vill be replaced by a fail open, DC rolenoid operated-flow control valve to control SG level. AF 3872 vill be normally open and have its'handvheel and local control station locked, with operation from the main control room'and SFRCS actuation signal retained. SG 1-2 Isolation Valve AF 599 sill'be normally open with the SFRCS actuation signal eliminated and auxiliary feedvater flow to the steam generator vill be limited to a maximum of 800 gpm by a cavitating venturi. A reviev' concluded that the continued flow of. auxiliary feedvater to a faulted steam generator at 800 gpm for a duration of 10 minutes vill not challenge containment pressure or actuate the-Containment Spray System. In addition, an evaluation of.the probability of.
recriticality with continued flow of auxiliary feedvater to a faulted steam generator at 800 gpm for a duration of 10 minutes and has concluded that the
, reactor vill not return to criticality under these conditions.
Changing Valve AF 3872 from normally closed to normally open vill ensure an open flovpath from AFF 1-2 to SG l-2 in the event of a loss of main feedvater.
The SFRCS close signal to AF 3872 is retained to isolate SG l-2 in the event of a main steam line break. Also, the steam generator can be isolated with AC povered AF 599 or DC Solenoid Operated Flow Control Valve AF 6451 by manual action if required. Backflow from SG 1-2 to either auxiliary feedvater pump is prevented by all possible backflow paths being provided with three check valves installed in series.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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n-SAFETY EVALUATION
SUMMARY
FOR' FCR.84-0209 TITLE:
". Fire Barrier at. Vest Vall of Room 516 CHANGE:
Install.tvo fire dampers (FD-1193 and FD-1194, reroute three conduits and revork the penetration seals in the vest vall of Room 516.
~ REASON FOR CHANGE:
- To upgrade the vest vall of Room 516 to a three-hour fire rated barrier SAFETY EVALUATION
SUMMARY
Fire Damper-FD-1193 vill be installed in the Auxiliary Building Non-Radvaste
, Area Ventilation System.. Because this system has no safety function the addition of a fire damper and security barrier cannot adversely affect.the' safety of the plant. based on an HVAC system analysis.
Installation _of penetration seals has no functional impact on the penetration.
These penetration seals are, required to maintain barrier integrity. This
- barrier is also a~ radiation barrier; therefore, seals meeting the. radiation requirements (grout, silicone elastomer) are specified.
None'of the three conduits to be rerouted are tied to'a system serving a safety function so" their determination and retenaination vill not adversely affect plant. safety.-
As summarized.above,.the proposed action vill not-increase the probability.or consequence of an accident or malfunction'previously evaluated in the USAR.
The~ proposed-action vill not create the possibility for an accident or n malfunction of.a different type than'any evaluated previously in"the USAR, and does.not reduce any margin of safety as defined'in the Technical
^ Specifications..
Therefore, an unreviewed safety. question does no? exist.
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SAFETY EVALUATION
SUMMARY
FOR R FCR'85-0065, Rev. O, Supp. 5 (SE 87-0202 Rev. 2) I
. TITLE: i Align AF-3870 Open i
L ' CHANGE: I Align Auxiliary Feed Fump 1-1 Discharge Valve AF-3870 from normally' closed to n'rmally o open, lock the handwheel and install a cover plate on the local control station to preveat inadvertent closure of the valve.
REASON FOR CHANGE:
i Aligning AF-3870 normally'open together with the modifications implemented by l FCR 85-0154, FCR 84-0190 and FCR 86-0330 vill improve the reliability of the Auxiliary Feedvater System.
I' ' SAFETY EVALUATION
SUMMARY
After implementation of FCR 85-0154, FCR 86-0330, and FCR 85-0065, with the plant in normal operation, the following condition vill exist: AFF 1-1 Discharge Valve AF 360 vill be replaced by a fail open, DC solenoid operated
, . flow control valve to control SG 1evel. AF 3870 vill be normally open and l have its handwheel and local control station locked with operation from the main control room and SFRCS actuation signal retained. SG l-1 Isolation ~ Valve AF 608 vill be normally open with the SFRCS actuation signal eliminated and auxiliary feedvater flow to the steam generator vill be limited to a maximum of.800 gpm by a cavitating venturi. A review concluded that the continued
. flow of auxiliary.feedvater to a faulted steam generator at 800 gpm for a
' duration of 10 minutes will not challenge containment pressure or actuate the Containment Spray System. In addition an evaluation of the probability of recriticality with continued flow of auxiliary feedvater to a faulted steam generator at 800 gpm for a duration of'10 minutes and has concluded that the reactor vill not return to criticality under these conditions.
L Changing valve AF 3870 from normally closed to normally open vill ensure an l open flovpath from AFF 1-1 to SG 1-1 in the event of a loss of main feedvater.
The SFRCS close signal to AF 3870 is retained to isolate SG 1-1 in the event of a main steam line break. Also, the steam generator can be isolated with AC powered AF 608 or DC Solenoid Operated Flow Control Valve AF6452 by manual action if required. Backflow from SG l-1 to either auxiliary feedvater pump is prevented by all possible backflow paths being provided with three check I' valves installed in series.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or l malfunction of a different type than any evaluated previously in the USAR, and does.not reduce any margin of safety as defined in the 'iechnical Specifications.
l i
Therefore, an unreviewed safety question does not exist.
' If SAFETY EVALUATION
SUMMARY
FOR FCR 85-0091'(SE 87-0198)
TITLE:
l-
. Control' Room Dust Control CHANGE :
Improve the' filtration capabilities of the Control Room Air Handling Units by
- 1) replacing the roll filters with disposable cartridge filters 2) increasing fan speed, 3) replacing the 15 HP fan motors with 25 HP fan motors and
- 4) replacing the differential pressure gauges on the filters.
REASON FOR CHANGE:
The roll filters in the Control Room Air Handling Units (S10-1 & S10-2) allow too much dust to pass through to'the Control Room. It has been identified that a build-up of dust could degrade equipment operation and safety system performance. The purpose of FCR Number 85-0091 is to improve the filtration capabilities of the air handling units and reduce the airborne' dust' build-up in the control room.
SAFETY EVALUATION
SUMMARY
The Control Room Normal HVAC System serves no safety function. The safety related functions of automatically shutting down the normal supply and return fans and closing the_ control _ room isolation dampers are not affected.
Initiation of these automatic actuations is on receipt of a SFAS signal or by high radiation ~or chlorine concentrations detected in the intake' air. The safety-related radiation and chlorine detection systems and their associated actuation circuitry are unaffected by this FCR as well as the circuitry associated with:the initiation of the SFAS signal.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an. accident or malfunction of a different type'than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR FCR 85-0096, Re5 B (SE 87-0318)
TITLE:
Redundant Steam Generator Level and Pressure Indication CHANCE:
Provide redundant pressure and level indication for both steam generators in the new center console in the Control Room.
REASON FOR CHANGE:
This addition vill allow the operators to analyze instrument response to steam
" generator transients and thus provide more reliable indication of steam generator conditions.
SAFETY EVALUATION
SUMMARY
Modifications made under this FCR involve indication only and do not affect automatic control functions. There is no adverse impact on the safety of the plant.
Steam Generator Level Indication The-signals to the new level indicators are isolated from the level transmitters and associated circuits by safety-related i. -tion devices.
The isolation devices are seismically qualified. Therefore, the safety function of the SFRCS vill not be affected as the isolation device vill prevent any indicator circuit malfunction from affecting the SFRCS circuits.
Steam Generator Pressure Indication The existing pressure ir.dicators and associated circuits which are being
, relocated are safety-related. The new installation of these existing indicators vill also be safety-related and vill comply with all criteria for the existing installation. Therefore, the function of the existing indicators vill not be affected by these modifications.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
i SAFETY EVALUATION
SUMMARY
FOR .
FCR 85-0109 (SE 88-0082)
TITLE:-
Install New Cabinet in Control' Room
, CHANGE:
c L Install a new cabinet in the' control room to facilitate the installation of new indications.
REASON FOR CHANGE:
The new panel is required in order to support the modifications being made to
.the startup feedvater pump, redundant SG level and pressure indications and the SFRCS enhancements resulting from the Detailed Control Room Design Review (DCRDR).
The centralizing and grouping of SFRCS and Auxiliary Feedvater System I
components should improve plant safety.by allowing the operator to rapidly-verify' proper automatic actuation and easily recognize any system problems.
SAFETY EVALUATION
SUMMARY
The relocation of.the controls and. indications performed by this FCR vill not affect the function or proper operation of the components. The function of the components vill not change because the reconnection vill be made in the
- same circuit. -The proper operation of the components will not be adversely .j affected since'the existing viring or equivalent standard cabinet viring vill i be tecminated at-the same termination points currently used.
The center console components installed by this FCR vill not impact plant safety.. The existing components vill be replaced with an identical, equivalent, or upgraded model. All components vill be purchased using "0" specifications.
Spacing of viring and ecmponents in the consoles vill be installed such that physical separation of 12 inches is maintained between redundant essential channels or metal barriers vill be utilized for maintaining separation.
Channel separation vill be maintained by routing Channel A cables with Channel I cables, and Channel B cables with Channel 2. .
i consoles C5702-C5714 and the console controls and indications will be seismically qualified.
As summarized above,'the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
the proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications. l
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Therefore, an unreviewed safety question does not exist. ]
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SAFETY EVALUATION
SUMMARY
FOR FCR 85-0124 TITLE:
Upgrade Steam Vent Mixing Condenser CHANGE:
Replace the existing Auxiliary Boiler Condensate Receiver (ABCR) quench or mixing type condenser with a shell and tube condenser.
REASON FOR CHANGE: j To overcome a design deficiency that resulted in contamination of the auxiliary boiler feedvater.
SAFETY EVALUATION
SUMMARY
The new ABCR condenser vill be cooled with the Turbine Plant Cooling Water System (TPCW). The TPCW does not fulfill any safety functions.
Lines leading from the TPCW header to the new ABCR heat exchanger and from the heat 2xchanger to the low level cooling water tank, would have a maximum flow in the range of 25 gpm, which is far smaller than the flow in the TPCV system.
Hence it can be concluded that no failure resulting from implementation of FCR 85-124'vould impact plant' safety.
In the unlikely event such a tube to shell leak should occur, it is not considered that introduction of TPCV, which is inhibited demineralized water, into the Auxiliary Boiler System would damage the auxiliary boiler.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
SAFETY EVALUATION
SUMMARY
FOR FCR 85-0154 REV. A, SUPP. 2 (SE 88-0434)
TITLE:
SFRCS Signals to AF-599 and AF-608 CHANGE:
Delete SFRCS safety-actuated signals from the control circuits for Auxiliary Feedvater Isolation valves AF-599 and AF-608 and place the valves in the open position. A blue light on the main control panel provides indication that the valve actuation circuitry has been disabled.
REASON FOR CHANGE:
Spurious closure of valves AF-599 and AF-608 by an inadvertent operator action and a subsequent failure to re-open vould completely isolate feedvater to both steam generators. This change vill remove the SFRCS signals to the valves and provide a means of removing control power once the valves are open.
SAFETY EVALUATION
SUMMARY
The safety function of auxiliary feedvater isolation valves AF 599, AF 3872, AF 608, and AF 3870 is to provide a feedvater path following an AFV actuation and to isolate the affected steam generator upon a SFRCS low steam generator pressure actuation (MSLB). Valves AF 599 and AF 608 are normally open, while valves AF 3872 and AF 3870 are currently in the closed position. FCRs 85-0065 and 84-0190 vill realign valves AF 3870 and AF 3872, respectively, to be normally open.
The effect of continuous auxiliary feedvater flow following a main steam line break on the containment temperature and pressure analysis was reviewed. With the installation of cavitating venturis by FCR 86-0330, the Auxiliary Feedvater flow into containment following a Main Steam Line break is limited l to 800 gpm. This modification vill eliminate the SFRCS signals to valves AF 599 and 608 and, thus, requires operator action to close the valves. Based upon a recent review, the continued flow of AFV to a ruptured steam generator at 800 gpm for 10 minutes vill not challenge containment pressure or actuate the Containment Spray system (the ten minute duration is imposed by the requirement for operator action to terminate the Auxiliacy Feedvater flow).
Therefore, this modification vill not affect the USAR analysis.
1 Refer to the safety evaluation summaries for FCRs 84-0190, 85-0065, and 1 86-0330.
This modification does not adversely affect the safety function of the Auxiliary Feedvater System and minimizes the probability of loss of all feedvater to a steam generator.
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- 1. As. summarized-above,-the~ proposed. action vill not'-increase the probability or-consequence of.an accident.or malfunction previously' evaluated in the USAR.
A' '
The proposed action vill not. create the possibility-for'an accident.or.
malfunction'of-'a'different type'than any evaluated previously in.the USAR,~and-Idoes n'tfreduce o any' margin of safetyfas defined in the Technical'
- 4. Specifications.
Therefore _ mi unreviewed safety. question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR FCR 85-0157 REV. A, SUPP. 01 (SE 88-0578)
TITLE:
Lover SFRCS SG Lov and High Level Trip Setpoints CHANGES-
- 1) Lower the SFRCS SG Low Level trip setpoint from 20" to 16.4" above the lower tube sheet. (License Amend 118)
-225" for'SU B.
REASON FOR CHANGE:
- 1) Lovering the low level trip setpoint This change vill minimize inadvertent SFRCS actuations by increasing the margin between the SFRCS Low Level Trip setpoint and the Integrated Control System (ICS) Low Level Limit, thereby improving Main Feedvater availability.
- 2) Lovering the high level trip satooint The high level setpoint should provide assurance that the setpoint will be effective in providing automatic termination of main feedvater in the
-event of an overfill condition without significantly increasing the possibility of spurious high level SERCS trips during plant operation.
The new analysis and setpoints reflect the dynamics of the Davis-Besse steam generators to the actual flow conditions.
SAFETY EVALUATION
SUMMARY
- 1) Lovering the low level trip setpoint The performance of the AFW System is described in the USAR Chapter 15 accident analyses for the Loss of Feedvater transient (Section 15.2.8).
This transient puts more severe design requirements on the AFW system than other transients and is the only design basis accident that takes credit for SFRCS due to lov steam generator level. The revised USAR analysis used an AFW flovrate cf 600 gpm to be delivered within 40 seconds of actuation of the AFUS.
- 2) Lovering the high level trip setpoint The function of the SFRCS high level trip is to minimize the potential for moisture carryover into the main steam lines. The selection of the high level setpoint should provide assurance that the setpoint vill be effective in providing automatic termination of main feedvater in the event of an overfill condition without significantly increasing the possibility of spurious high level SFRCS trips during normal plant operation.
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As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not. create.the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and v does not reduce any margin of safety e: defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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~ SAFETY EVALUATION
SUMMARY
FOR FCR 85-0198, REV. B E '
TITLE:
CVRT Permanent! Filter Installation 1
- l l! ' CHANGE:
Frovide for the permanent. installation'of a filter to remove particulate matter from liquid stored ~in the Clean Waste Receiver Tank (CVRT). 'i l' REASON FOR CHANGE:'-
l
. To lover' radiation levels in the CVRT Rooms and improve water: chemistry in the CVRTs.
i SAFETY EVALUATION
SUMMARY
The filter unit that is being permanently installed by this.FCR has=been used' y on a temporary baris to provide a clean-up function for the CVRTs. The l
installation of the permanent filter: vill not decrease the integrity.of the clean vaste system.
As summarized above,Lthe proposed action vill notLincrease the probability or consequence of an accident or malfunction previously evaluated in the USAR.
' The' proposed action vill not create the possib:lity for an accident or
. malfunction lof a'different type tiian'any evaluated previously'in the USAR, and
~
i does not reduce any margin of safety as defined'in the' Technical Specifications.
/ Therefore, an.unreviewed safety question does not exist.
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" (i SAFETY EVALUATION
SUMMARY
FOR j FCR 85-0207 REV. 1 (SE 88-0006) l LTITLE:
CRDM Ventilation System j l
-CHANGE:
Provide a.new CRDM ductvork ven*ilation system including two new centrifugal l
' fans. l t
REASON FOR CHANGE:
l To' eliminate the repeated failures which occurred with the six vertically l meunted vaneaxial fans originally. installed in the CRDM Ventilation System.
SAFETY EVALUATION
SUMMARY
. .i i
LAlthough the CRDM ventilation system is not safety-related, the entire CRDM -l ventilation system is designed.and installed to Seismic Category I. I requirements to preclude any Seismic II/I hazards.- The permanent ductwork,
' including the removable duct sectionibeing stored, meets the Seismic ;
Category I requirements for Seismic II/I concerns during refueling. l l
The materials and paint used for.the ductvork, supports and fans do not l contain agents which would contribute to additional hydrogen generation inside j the containment. ,
The CRDM service structure ventilation fans are considered incapable of j ger:erating missiles dangerous to essential equipment. l I
As summarized above,.the proposed action vill not increase the probability or ;
consequence of an accident or malfunction previously evaluated in the USAR. !
The proposed action vill not create the possibility for an accident or ,
malfunction of a different type than any evaluated previously in the USAR, and '
does not reduce any margic of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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s i j SAFETY EVALUATION
SUMMARY
-FOR FCR 85-0221-03 (SE 88-0512)
TITLE:
Modifications to Service Vater Lines for the Control Room Emergency Ventilation' System (CREVS) Condenser
' CHANGE'
0 2. Change status of PDI 9808 and PDI 9809 from temporary status to permanently. installed.
REASON FOR CHANGE:
SV 132 and SV 136 vill-be maintained locked open to prevent inadvertent closure and isolation lof a standby system.
PDI 9809 and PDI 9809 provide valuable performance information during monthly and refueling surveillance testing and are used as a basis to ensure the flow control valve is' operating properly.
SAFETY EVALUATION
SUMMARY
Changing the position of valves SW 132 and SW 136 from locked throttled to
' locked open vill'not have any effect on CREVS operation as the flow is nov
' throttled by FCV 5896 and FCV 5897.
The change of PDI 9808 and PDI 9809 to permanent status vill not have any adverse effect on the safety of the plant.
l- As summarized above, the proposed action vill not increase the probability or consequence of'an accident or malfunction previously evaluated in the USAR.
The' proposed action vill not create the possibility for an accident or malfunction of a different' type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
l .I
SAFETY EVALUATION
SUMMARY
FOR FCR 86-0036 (SE 88-0408, Rev. 1)
TITLE:
RCS Extended Range Pressure Indicator CHANGE:
Install two pressure transmitters for RCS extended range pressure indication.
Indications vill be provided on the post accident panels in the main control room and the auxiliary shutdown panel.
REASON FOR CHANGE:
To comply with Regulatory Guide 1.97 which recommends providing two channels of 0-3000 psig pressure instrumentation for the RCS.
SAFETY EVALUATION
SUMMARY
This modification does not decrease the reliability of the RCS. Design, materials, and installation vill comply with the applicable requirements of the Quality Assurance Manual. Additionally, all installation vill meet Seismic Class I standards; no RCS pressure boundary vill be degraded.
The function of the transmitters is to provide RCS pressure indication for transients in which the RCS pressure exceeds 2500 psig. The indication can be utilized to verify that the automatic protective actuations successfully mitigate the consequences of the transient.
As summarized above, the proposed action vill not increase the probability or
' consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR FCR 86-0123 (SE 87-0301)
TITLE:
Reactor Power Auctioneer to the Integrated Control System (ICS)
. CHANGE:
Modify the auctioneering circuitry which provides the power range Nuclear Instrumentation signal to the ICS.
This modification consists of:
a removing averaging circuitry from RPS Channels B&C;
- relocating Power Range Nuclear Instrumentation test switches before the first stage of auctioneering circuitry;
- installing a first stage of auctioneering circuitry in each train of Non-Nuclear Instrumentation; and
- installing a second stage of auctioneering in the ICS.
REASON FOR CHANGE:
To eliminate an existing fault of the circuitry in that a loss of channel power to an RPS Channel coincident with one channel of Nuclear Instrumentation being in " test" may result in loss of the power range signal to the ICS, and a subsequent plant trip.
SAFETY EVALUATION
SUMMARY
The ICS is not required for safe shutdown of the plant. This modification vill reduce the probability that a failure of ICS input (i.e., neutron power) vill occur, and thus reduce the frequency of challenges to plant safety systems that result from ICS perturbations.
As summarized above, the proposed actions vill not increase the probability or consequence of an accident or malfunction previousIy evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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n-SAFETY EVALUATION
SUMMARY
FOR FCR 86-0192, SUPP 2 (SE 87-0263)
TITLE -
d Repower the Startup Feedvater Pump CHANGE:
i Repover the Startup'Feedvater Pump (SUFP) and reconnect the' associated piping systems.
REASON FOR CHANGE:
.The reason for'this modification 3s to return the SUFP to an operable-status.
A This FCR vill: accomplish the following to return the-SUFP to an operable status:
-' The SUFP vill be repovered from 4.16KV Bus C2.
- -- DThe handvheelifor valve FW32 vill be reinstalled, but the valve vill be locked closed, to provide a means to manually align the SUFP suction to the Demerator Storage Tanks (DSTs).
- .- The SUFP-minimum flow recirculation line to the Demerator will be reconnected.
Valves CW134 and CW135 which provide the SUFP seal water cooler and lube oil cooler return paths, respectively, vill be positioned normally open.
Valves CW196 and CW197 which isolate TPCW outside the AFV pump rooms vill be locked closed.
- Valves FW500B and FW495 vill be positioned normally open to align SUPP suction and discharge pressure indication, respectively.
A new SUFP local indication panel vill be installed in Room 238'for monitorice the-SUFP pump / motor bearings and stator temperature indications. A new light to indicate excessive vibration of the SUFP will also be mounted on the new panel. A horn activated by the vibration monitor switch will be installed in Room 238.
l' ' SAFETY EVALUATION
SUMMARY
Returning the SUFP system to an operable status vill provide a means to remove decay heat via the steam generators in the unlikely event that all other feedvater sources are lost. The SUFP system was part of the original plant design and was removed from service after it was discovered that unanalyzed piping failures could jeopardize the AFV r. amps. The SUFP system serves no safety function.
1
_--.--_x--.a.-- - - -_-----,--u.--,.. . x.- ------.._--.a.- - - _ - - - - - . - - - -----,---._-a---
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l 1 ,
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-The AFV system vill not be.'advursely affected by reterning the SUFP system to an operable' status because of.the i'ollowing:
- 1. The SUFP suction and discharge lines v 4 the TPCW lines.in the AFW rooms vill continue to be isolated by means'of manual valves located'outside of the AFV pump rooms.
2.-~FCR'85-0296 removed the handvheels from valve FW32 ar.d FW91 to reduce the .
possibility of the AFV pumps; inadvertently being aligned to take suction from the DSTs. The concern vas that'the high temperature DST vater could damage the AFW pump bearings if the DST was used as the suction source.
Since valve FW91 vill reme.in positioned normally. closed with its handwbeel removed,-reinstalling the handwheel on valve FW32 vill not. increase the possibility of inadvertently aligning the AFV pumps to take' suction from the. DST. . FCR 84-021 positioned valve FW32 to be'nornally closed.
- 3. The SUFP was installed in Raom 238 (AFWP l-2) as part of the original plant design.' The design for the ventilation system in this room included
, the SUFP system heat load.
As_ summarized above, the proposed actio'n. vill not increase the probability or-consequence of an accident- or malfunction previously evaluated in the. USAR.
The proposed action vill not create,the possibility for an accident or malfunction of a'different type than'any evaluated previously in the USAR, and:
does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed. safety question does'not exist.
1 2
SAFETY EVALUATION
SUMMARY
FOR FCR 86-0220A TITLE:
Upgrade Fire Dampers CHANGE:
Replace Fire Campers FD-1076, FD-1077, FD-1084, FD-1093, and FD-1166; remove Fire Dampers FD-1101 and FD-1102.
REASON FOR CHANGE:
- 1. Fire Dampers FD-1076, FD-1077, FD-1084, FD-1093 and FD-1166 vere not originally installed per manufacturer's and NFPA Code requirements.
- 2. Fire Damper FD-1101 is not accessible for rework due to high radiation levels in the Spent Resin Tank Room.
- 3. Fire Damper FD-1102 is no longer required as a result of the Fire Area Optimization Frogram.
SAFETY EVALUATION
SUMMARY
Modification of the fire dampees installed in the Radvaste Area Ventilation System vill not adversely affect the safety of the plant based on an HVAC system analysis because the system does not perform a safety function. The fire dampers do perform a safety function by virtue of their presence in fire barriers protecting redundant trains of safe shutdown (App. R) equipment.
Fire Damper FD-1101 is not accessible for rework due to high radiation levels on the Spent Resin Tank Room (102) side and extensive interferences on the Corridor (209) side. The existing fire damper vill be abandoned in place sinco access for removal is impossible. Fire Damper FD-1102 vill be removed as part of the Fire Area Optimization Frogram. Based on the results of the Fire Area Optimization Report the two fire areas currently being separated by FD-1102 vill be combined into one fire area. Because of this combination FD-1102 is no longer required. Safe shutdown (App. R) is still assured after the combination of Fire Area's I and G so removal of FD-1102 does not adversely affect plant safety.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist. I
-= . _ _ _
SAFETY EVALUATION
SUMMARY
FOR FCR.86-0226A I
TITLE:
Upgrade Fire Dampers CHANGE:
- 1. Change out eight fire dampers (FD-1040, FD-1041, FD-1043, FD-1046, FD-1047, FD-1060, FD-1061 and FD-1163)
- 2. Remove three fire dampers (FD-1051, FD-1112 and FD-1113)
REA60N FOR CHANGE:
- 1. The eight dampers being changed vere not originally installed per manufacturer's and NTFA Code requirements.
- 2. FD-1051 is being removed because it is not installed in a fire barrier.
- 3. FD-1112 and FD-1113 are being removed because they are installed at the end of a duct run which terminates at fire barriers which are protected by automatic roll up fire doors.
SAFETY EVALUATION
SUMMARY
The function ni the Auxiliary Building Non-Radvaste Area Ventilation System is to provide a suitable environment for equipment and personnel by maintaining the area temperatures between 60 and 120 degrees Fahrenheit. Modification of the installed fire dampers vill not adversely affect the safety of the plant based ,n a HVAC system analysis because the ventilation system does not perform a safety function.
The function of Appendix R fire barriers is to contain a fire within one specific area, thereby, ensuring the availability of one train of safe shutdown equipment. The fire dampers are installed in HVAC ductvork when a section of ductvork penetrates a fire barrier so the integrity of the fire barrier is maintained.
The fire dampers are being replaced by equivalent fire dampers that will be installed in accordance with a tested configuration. The three fire dampers to be removed are not required to perform a safety function and are not installed in a system required for accident mitigation.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction prevfously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
SAFETY EVALUATION
SUMMARY
FOR FCR 86-0291 TITLE:
Modify High Pressure Injection System Flow Test Line l CHANGE:
l Hodify the High Pressure Injection (HPI) System Flow Test line to allow for flow testing of the HPI pumps at higher flow rates (greater than 600 GPM).
REASON FOR CHANGE:
An increase in testing flow is desired to allow for more HPI pump testing flexibility, to reduce the impact of test instrument inaccuracies, and to provide a more acceptable demonstration of pump operability.
SAFETY EVALUATION
SUMMARY
The purpose of this FCR is to enhance pump testing operation to provide a better indication of pump operability. The accident mitigatin e 'eatures of the HPI system vill not be affected by this change. Administrate .e controls are in effect during testing that require operator action to restore the inoperable (tested) train to service if it is needed to perform a safety function during testing.
The modifications performed for this FCR involve passive components that are being installed in accordance with Seismic Category 1 requirements, and vill be nuclear safety-related where required. Components that are not safety-related vill be normally isolated by a closed valve, serving as the function 0 boundary.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
E h
g 1
SAFETY EVALUATION
SUMMARY
l
.FOR l FCR 86-0318 (SE 87-0239) .i q
TITLE: '
l Relocate Hot Leg Level Monitoring System Reference Leg CHANGE:
i Relocate the Hot Leg Level. Monitoring System (HLLMS) reference leg pipe / tubing j
- tie-in point to the top of the hot leg vent pipe for both HLLMS channels.
REASON FOR' CHANGE: ,
'i To eliminate the potential adverse effect the hot leg vent line could have on
~
'I
' the HLLMS indication due to the presence of non condensable gases which would !
prevent the complete refilling of the reference leg.
- l. i
' SAFETY EVALUATION
SUMMARY
Although.the'HLLMS serves no safety function this modification.should minimize the HLLMS measurement error resulting after a hot leg venting evolution. '!
All new piping and tubing vill be installed to Seismic' Category I requirements. The new 3" CCA-20 & 1/2" CCB-36 piping vill meet ASME Section-III, Class 1 and 2 requirements, respectively. Since all lines added.
i do not exceed 1" nominal diameter, high energy line breaks are not postulated !
to occur. The ASME Section III piping and valves have been added to the l Davis-Besse ISI program. A piping system hydro vill verify RCS pressure boundary integrity. )
1
' Reference leg routing vill maintain the original HLLMS separation criteria of j,
, routing tubing within the respective steam generator compartment. HLLMS ;
instrument isolation and separation are not affected by this FCR. !
- As summarized above, the proposed action vill not increase the probability or ;
~
consequence of an accident or malfunction previously evaluated in the USAR. !
The proposed action vill not create the possibility for an accident or )
malfunction of a different type than any evaluated previously in the USAR, and l does not' reduce any margin of safety as defined in the Technical i Specifications.
. Therefore, an unreviewed safety-question does not exist.
i J
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SAFETY EVALUATION
SUMMARY
FOR FCR 86-0330, REV. B (SE 87-0292 REV. 2)
TITLE:
Auxiliary'Feedvater Level Control CHANGE:
.1. Eliminate the SG leval control system utilizing automatic variable speed control of the Auxiliary Feedvater Pump Turbine. (AFPT)~and replace it with modulating solenoid flow control valves. These valves vill be installed in place of AF 360 and AF 388.
- 2. Install cavitating venturis in the AFV lines to the steam generators.
REASON'FOR CHANGE:
- 1. Operating experience has shown.that the existing level control scheme to be unstable with AFit speed continuously cycling to and from the high and-lov speed stops (3600 rpm and 1950 rpm respectively) and AFV flow cycling from 0 to greater than 1000 gpm. .This on/off cycling of AFV flow results in undesirable steam generator level and pressure oscillations as well as a large number of thermal cycles on the AFW external' header.
- 2. The venturis are being installed to limit AFV flow to each steam generator to e maximum of 800 gpm.
SAFETY EVALUATION
SUMMARY
This modification removes normally open valves AF 360 and AF 388 and replaces them with DC povered modulating solenoid flow control valves. The new flow control valve is designed to fail open on loss of power. This is'an improvement over the existing design in that the valve failing open vill facilitate continued AFV flow. The valves will be interlocked with the AFPT steam
. admission valve (MS 5889 A or B),so that it will remain open and not receive control signals during normal operation. On an SFRCS initiation signal, the AFPT steam admission valve vill oper and the AFPT will accelerate to 3600 RPM.
The AFPT speed will be held const:at at 3600 RPM by the governor valve. As the steam admission valve begins to open, control signals will be supplied to the flow control valves from a new Foxboro control system. The Foxboro control system vill be housed in two new cabinets which vill provide channel separation for both AFPT flow control systems.
The capability vill be provided to manually position the flow control valves using hand / auto stations installed in the control rocm. The existing capability to manually control AFFT speed from the control room or the auxiliary shutdown panel vill be retained to allow flow control should the new flow control valve
. fail open (e.g., loss of power). AFPT speed would also be reduced to facilitate a plant cooldown using AFW.
This modification also installs a cavitating venturi in each AFV line to limit the' maximum flow rate to each steam generator to 800 gpm regardless of steam generator pressure. Calculations verify that AFV and motor driven feed pump flows following implementation of this modification vill still exceed the 600 gpm AFV flow requirement specified in the bases section of Technical Specification 3/4 7.1.2.
The impact of limiting AFV flow on the SBLOCA analysis which assumed greater than 800 gpm flov have been evaluated and concluded to have no adverse impact.
Refer to the safety evaluation summaries for FCRs 84-0190, 85-0065, and 85-0154 for additional changes to AFV system.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR. The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore. an unreviewed safety question does not exist.
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1 iT SAFETY EVALUATION
SUMMARY
FOR' 3 FCR 86-0334,-SUPP. 2 O
cg LTITLF;
' Modify Station Air Compressor Pressure Safety Valve PSV 2119
' CHANGE;
^ . , ,
sc hange:the velded-connection to a, flanged connection for Station Air Compressor Relief Valve PSV.2119.
, REASON FDR CHANGE:
.To; facilitate removal for maintenance and repair.
' SAFETY' EVALUATION
SUMMARY
Valve PSV 2119 has'no effect on plant safety. By revelding the connection to the'PSV every time it'is serviced, the steel is'veakened and.the potential for failure increases. In this case'the flanged connection is more reliable and requires less maintenance than a velded connection.
' As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or',
malfunction of a different type than any evaluated previously.in the USAR, and does not' reduce any margin of safety.as defined in the Technical' Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR FCR 86-0425, SUPP. 8, REV. 2 (SE 88-0587, REV. 1)
TITLE:
Motor Driven Feedvater Pump (MDFP) Phase II CHANGE:
1 This modification providee varicus improvements to the HDFP system that was added und.er FCR 85-025. These changes include relocating all local HDFP '
controls and instrumentation to a common panel, adding a service water supply to the MDFP suction, adding two solenoid operated flow control valves to each AFV line, removing MDFP discharge motor operated valve FV 5867, rerouting the test line, rerouting the varmup line, adding restriction devices in the first stage cooling water outlet lines, adding vents to coolers, adding position !
switches to valves, and revising the position of valve SV 147 to locked open.
REASON FOR CHANGL: j These improvements are being made to enhance the reliability of the MDFP system to supply feedvater to the steam generators.
SAFETY EVALUATION
SUMMARY
The MDFP System does not perform a safety function and is not required to operate during emergency shutdown. It is, however, used as a backup to the Main and Auxiliary Feedvater systems for supplying water to the Steam Generators in case either of these two systems are lost. In addition, the MDFP is designed to provide a continuous feedvater supply to the Steam Generators during plant startup and normal shutdown.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accider.t or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
F SAFETY. EVALUATION
SUMMARY
FOR-FCR 86-0432, SUPP. 7 (SE 88-0219 REV. 1)
TITLE:
Enhanced Feed and Bleed Capability CHANGE:
Upgrade the' plant's feed and bleed cooling capability by. modifying the Pilot Operated Relief Valve (PORV), Ma'keup System (MU), and supporting auxiliaries.
The following modifications will be performed:
- The MU pump suction from the BVST will be relocated to the HPI suction.
lines. This modification vill allow piggyback operation of the MU System on the LPI pumps in order to increase flowrate.
-- To further increase injection flow to-the RCS during feed and bleed cooling, solenoid valves will be installed on the MU pump minimum recirculation flov lines to permit manual isolation.
- The SFAS signal on MU33 vill be deleted. This vill prevent deadheading the pumps when SFAS is initiated during feed and bleed cooling. 'The deletion of this signal from HU33 requires a Technical Specification change (Amm 112) which is being pursued under FCR 87-0131.
MU211, the bypass valve around normal control valve MU32, vill be replaced with a motor operated valve (MOV). This MOV vill be opened during feed and bleed to lover system flow resistance.
- The existing MU pump discharge piping ties into HPI Train 2. A second parallel injection path will be provided to HPI Train 1.
- Various support systems vill be modified including provision for essential CCW supply to the MU pump oil coolers, replacement of MU pump lube oil pump motors and switching'the PORV power supply to essential DC power.
REASON FOR CHANGE:
It has been shovn by analysis that the existing plant design can sucetssfully accomplish feed and bleed cooling provided both makeup pumps and the PORV are l opetable. However, Toledo Edison has committed to enhance the current' feed and bleed capability. Feed and bleed is not a part of the design bisis of Davis-Besse and this upgrade represents an enhancement of the post-trip core cooling capacity.
.___-__________-_-____ _ A
o I
1he upgraded system vill provide sufficient feed and bleed cooling to ensure that the liquid level in the reactor vessel maintains the core covered.
Successful; feed and bleed cooling would be provided by either of the following
- configurations:
I-A. Two mak.eup pumps and the pressurizer code safety valves B. One makeup pump-in " piggyback" with the LPI System and the PORV I
SAFETY EVALUATION
SUMMARY
l 1
The objective of this-cooling method is to provide MU feed'and bleed cooling I . either until secondary side cooling is restored or the plant is depressurized to below 1600 psi to allow feed and bleed cooling through the HPI System. The enhanced makeup system provides additional redundancy and reliability to ensure core cooling following an off-design basis loss of all feedvater event.
The Makeup System is not taken credit for in any Design Basis Accident events and it has been verified that all modifications made under FCR will not adversely affect any required safety systems. The modifications discussed above vill enhance the plants response to an off-design basis total loss of feedvater event by enhancing the pr! mary side cooling capability.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
2
SAFETY EVALUATION
SUMMARY
FOR FCR 86-0432, SUPP. 8 (SE 88-0604)
TITLE:
Enhanced Feed and Bleed Capability (Close.MU 6421 and MU 6423B)
CHANGE:
Change valves MU 6421 and MU 6423B from normally open to normally closed.
REASON FOR CHANGE:
Supplements 0 through 7 of this FCR added an additional flow path for makeup to the RCS for enhanced Feed and Bleed capability. The purpose of this supplement is to implement corrective action from PCA0R 88-0576, Thermal Sleeve Failure, and NRC Bulletin 88-08 revising the feed and bleed design to isolate the flow path during normal operation.
SAFETY EVALUATION
SUMMARY
Closing valves MU 6421 and MU 6423B has no effect on the normal operation of the Makeup System. Both of these valves are in the second makeup injection path added in Supplements 0 thru 7 of this FCR. As described in the safety evaluation issued in Supplement 7 the new injection path is not intended for normal makeup operation.
Closing valve MU 6421 vill require additional operator action to actuate the makeup system in the feed and bleed mode. Valve MU 6421 vill need to be opened within 10 minutes of RCS hot leg temperature reaching 600*F. Valve MU 6421 may easily be opened within ten minutes using handsvitch HIS-6421.
Changing valve MU 6421 to normally closed does not adversely affect the Containment Isolation System. The remote manual isolation capability of valve MU 6421 is not affected by keeping the valve closed during normal operation.
l As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR FCR 87-0063, REV. 1 (SE 88-0031,.REV. 1)
TITLE:
Remove SFAS Actuation Signals & Relocate Control Switches CHANGE:
Remove the SFAS actuation logic signals from valves FW 601 & 612, MS 100, 101,
'375, 394, 100-1, & 101-1, and ICS 11A & 11B, and relocate their control switches from panel C5717 to the control room center console.
REASON FOR CHANGE:
The relocation of control switches in this FCR represents part of the disposition _of HED 9.2.001 and 9.2.033 identified during the Davis-Besse l
Detailed Control Room Design Review (DCRDR). HED 9.2.001 and 9.2.033 involve the centralizing and grouping of SFRCS and Auxiliary Fedvater System components, respectively. Deletion of Safety Features Actuation System (SFAS) signals to main steam and feedvater isolation valves was a recommendation from the Decay Heat Removal Task Force.
1 E SAFETY EVALUATION
SUMMARY
The relocation of the main feedvater isolation valve control switches will not affect'the' function or proper operation of the switches or valves. The
. function of the control switches will not change because the reconnection made by FCR 85-109 vill be in the same circuits. The proper operation of the switches and' valves will not be affected because either the disconnected cable or an=in-line. Class 1E termination vill be used for terminating in the new terminal box.
I Removing the SFAS. signals from these valves vill improve plant reliability by l reducing-the effect of inadvertent SFAS actuations which can cause a loss of l: ' main feedvater. The currently installed automatic SFAS signals are redundant
! vith the SFRCS lov steam Ifne pressure trip signals supplied to the affected valyes and.' the remote manual- operation capabili ty available for these valtes. l By utilizing the SFECS trip, and the actual existence of the pressute ;
gradients which eaist following a large LOCA, all the assumptions mtde in the l USAR accident analyses are met and the consequences are within the previously ]
analyzed bounds. ;
i As summarized above, the proposed action vill not increase the probability or ,
consequence of e.n accident or malfunction previously evaluated in the USAR. l i The proposed action vill not create the possibility for an accident or l L~
malfunction of a different type than.any evaluated previously in the USAR, and 1 does not reduce any margin of safety as defined in the Technical ,
Specifications.
Therefore, an unreviewed safety question does not exist.
I I
SAFETY EVALUATION
SUMMARY
l FOR FCR 87-0064 (SE 87-0306)
TITLE:
I Relocation of AFPT Isolation Valve Control Switches CHANGE:
' Relocate the Auxiliary Feedvater Pump Turbine (AFPT) isolation valve control switches for valves MS-106, 106A, MS-107, and MS-107A to the control room center console and add SFRCS block switches for these valves to permit !
overriding the SFRCS signal after automatic SFRCS actuation. l REASON FOR CHANGE:
-The relocation of control switches in this FCR represents part of the disposition of HED 9.2.001 and 9.2.033 identified during the Davis-Besse 4 Detailed Control Room Design Review (DCRDR). HED 9.2.001 and 9.2.033 involve 1 the centralizing and grouping of SFRCS and Auxiliary Feedvater System components,-respectively. Adding SFRCS block switches.to the AFPT steam supply isolation valves was a recommendation from the Decay Heat Removal Task Force.
SAFETY EVALUATION
SUMMARY
The' relocation of the AFPT steam isolation valve control switches will not affect the function or proper operation of the switches or valves.
The SFRCS block switches will not interfere with automatic SFRCS actuation or accomplishment of the safety functions. In accordance with IEEE-279, once initiated the SFRCS protective action at the system level vill go to completion and remain in the tripped state until deliberately reset by operator action. The SFRCS block switches will provide a manual override capability for the AFPT steam supply isolation valves to al3cv the operator to align steam to the AFPTs as he deems necessary after automatic SFRCS actuation.
As summarized above, the propcsed action vill not increasu the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
(4s SAFETY EVALUATION
SUMMARY
FOR .
FCR 87-0065 (SE 87-0369)~
' TITLE:
AFV Isolation Valves SFRCS Block Switches CHANGE:-
Add Steam and Feedvater Rupture Control System (SFRCS) block switches to the
-control room center console for valves AF 3869, 3870, 3871 and 3872.
REASON FOR CHANGE:
The SFRCS block switchesLwill' provide the operators with the ability to override the SFRCS signal after the completion of automatic actuation. Adding SFRCS' block switches to the auxiliary feedvater (AFW) isolation valves was a
~
recommendation from.the Decay Heat Removal Task Force.
SAFETY EVALUATION
SUMMARY
'The SFRCS block.svitches will not int'erfere with automatic SFRCS actuation or' accomplishment of~the above safety functions. In accordance.vith IEEE-279, once initiated the SFRCS protective action at the system level vill go to completion and remain in the tripped state until deliberately reset'by
. operator action. .The SFRCS block switches vill provide a manual override capability for the Auxiliary Feedvater Pump discharge valves to allow the operator.to control AFV flow to the steam generators as deemed necessary after automatic SFRCS actuation.
As summarized above, the proposed action vill not increase the probability or consequence of.an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of ssfety as deffned in the Technical Specifications.
Therefore, en unreviewed safety question does not exist.
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I' SAFETY EVALUATION
SUMMARY
FOR_
FCR 87-0067-(SE 87-0284)
-TITLE:
Control ~ Room' Center Console CHANGE:. ~
1
-Add valve position indication to the control room center console for the following> valves:
FV-SP-6A, D Hain Feedvater Control Valves FV-SP-7A, B Main Feedvater Startup Control Valves TSV-1, 2, 3, 4 Main Turbine Stop Valves REASON FOR CHANGE:
To address part of the concerns-identified by HED 9.2.001 and HED 9.2.054 that {
the present SFRCS display of.certain actuated components does not allow the j
. operator _to adequately verify _ proper SFRCS' actuation. t SAFETY EVALUATION
SUMMARY
The proposed modification vill not have any effect on plant safety. .This FCR
. merely routes signal ' cables for the subject valves to a new termination box which will be installed-in the cable spreading room. The valves are H physically located in the Turbine Building Area 5, at Elevation 603'. The valve position signal cables vill be routed in non-safety related conduit / cable trays to-the termination box.
Maintaining proper channel separation within the center console and performing the seismic analysis of the reconfigure console were disensed in the safety evaluation associated with FOR G5--0109. N As summarized above, the proposM action vill not herease the probability er consequence of an esecident or malfunction previously evaluatcd in the USAR. l k
The proposed action vill not create the possibility for an accident or j malfunction of a different type than any evaluated previously in the USAR, and j does not reduce any margin of safety as defined in the technical 'j Specifications.
Therefore, an unreviewed stafety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR FCR 87-0069 (SE 87-0207)
TITLE:
Auxiliary Feedvater Flow Indication CHANGE:
Provide a redundant escential channel of auxiliary feedvater (AFV) flov indication for each AFV train.
. REASON FOR CHANGE:
The addition of the redundant essential flow loop provides the instrumentation required to meet Regulatory Guide 1.97 for AFV flow indication SAFETY EVALUATION
SUMMARY
This modification vill not have any adverse effect on plant safety. The new flow loops.F-6426 and F-6427 provide flow loop channels redundant to existing flov loops F-4630 and F-4631. This redundancy will ensure that flow indication to each Steam Generator is available through.tvo loops, one fed by
' Channel 1 and the other by Channel 2, for control and operation of the station through all operating conditions, including anticipated operational occurrences, accidents and post accident conditions. Loop components will be environmentally and seismically qualified. Channel redundancy and separation vill be maintained in accordance with USAR requirements.
As summarized abova- the proposed action vill r.ot increase the probability or consequence of an ident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or ,
i malfunction of a different type than any eva.luated previoesly in the USAR, and "
does net reduce ar;y margin of safety as defined in the Technical Specificat!ons.
1herefore, an unreviewed safety question does not exist.
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a SAFETY EVALUATION
SUMMARY
FOR FCR 87-0070 (SE 87-03'90) i:
TITLE::
s Control Room Center Console l ~ CHANGE:
' Relocate'the following control switches from the feedvater panel'(C5721) to i
the control room center console (C5706 through C5710):-
.HIS-1382/1383 AFP 1/2 Service Water Suction Valve HIS-611/603 .SG 1/2 Blowdown System Isolation Valve HIS-611B/603B SFRCS Block Switch for Valve HV-611/603 i HIS-780/779 SG 1/2 Main Feedvater Isolation Valve ~
REASON FOR' CHANGE:
The relocation of control switches in this FCR represents part of the disposition.of HED 9.2.001 and 9.2.033 identified during the Davis-Besse Detailed Control Room Design Review (DCRDR). HED 9.2.001 and 9.2.033 involve.
the centralizing and grouping of SFRCS and Auxiliary Feedvater System components,'respectively.
SAFETY EVALUATION
SUMMARY
1The relocation'of the control switches vill have no effect on their safety functions or those of.any other components, systems or structures. This FCR
- simply disconnects the cables to these switches oa.the feedvater panel and then reroutes'and terminates them at the new terminal box in the cable spreading room.
The effects on safety due to the viring from the new terminal box to the l center' console and the reconfiguration cof the console is addressed in the
' safety ' evaluation for FCR 85-0?.09.
~As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
.The proposed tetion vill not: create the possibility for an' accident or malfunction of a different type than any evaluated previously in the USAR, and does not. reduce any margin of safety as defined in the Technical Specifications.~
Therefore, an unreviewed safety question does not exist.
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rs. -;p SAFETY' EVALUATION
SUMMARY
FOR FCR 87-0071 SUPP. 1 (SE 88-0461)
TITLES.
Relabel System Bypass Status Indicating Light IL/HS 4808 D ,- CHANGE:
Change manually operated,: System Bypass status indicating light IL/HS 4808 from "CTMT RECIRC" to " DIESEL CEN".
RJ3 SON F0it CHANGE:-
This label' change is required to provide the operators visual indication of the status of the Emergency: Diesel Generators, a nuclear safety related system.
SAFETY EVALUATION ~
SUMMARY
The proposed label change vill have no adverse effects.on safety because the Containment Recirculation System no longer requires a system bypass status indicator. FCR 84-0011 deleted the system's operational and surveillance requirements from the: Technical Specifications.
The nomenclature change of the indicating light and switch vill have no affect on the Safety Features Actuation System. IL/HS 4808 is isolated from the rest of.the system. In fact, by providing additional information on the status of' the Emergency Diesel Generators, this label change vill enhance present Safety-Related Display Information available to the operators.
A.1 summarized above, the proposed action vill hot 'lucrease the probability or consequence of an recident or'malfuncilen previously evaluated in the USAR.
.The proposed action vill not crerte the possibility for,an accident or malfunction of a different type than any evaluated previously in the USAR, and ;
does not reduce any margin of safety as defined in the Technical ~1 Specifications. !
Therefore, an unreviewed safety question does not exist.
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' SAFETY EVALUATION
SUMMARY
l FOR FCR 87-0092, REV. A (SE 87-0364) &
FCR 87-0130 (T.S. Amm 124) (SE 87-0261)
TITLE:
Modification of AFV/SFRCS Manual Initiation Switches l
CHANGE:
This FCR in conjunction with FCR 85-0109 & MOD 87-1107 relocates and reduces the number of SFRCS manual initiation switches. Specific modifications that are within the scope of this FCR are as follows:
- 1. Remove the ten existing SFRCS' manual initiation switches from feedvater control panel C 5721.
- 2. Install four new SFRCS manual initiation switches HIS-6401 and HIS-6403 for Channel 1.and HIS-6402 and HIS-6404 for Channel 2, on the new SFRCS panel being installed in the control room by FCR 85-0109.
REASON FOR CHANGE:
The desirability of improving the existing manual initiation scheme.vas illustrated during the Davis-Besse incident of June 9, 1985. Subsequent to this incident several reviews identified the need to improve the means of manual initiation. Specifically, FCRs 87-0092 & 87-0130 vill satisfy the concerns regarding manual initiation of auxiliary feedvater raised in Section II.C.5 of " Toledo Edison Company Davis-Besse Nuclear Fover Station Course of 4 Action Report" dated September 9, 1985; and provide a revised scheme for $FF.CS {
manual actuation. buttons, of Sections 3.3.3. and 3.3.4 of NUREG-1177, Davis-Besse Bestart SER.
, SAFETY UJALUATION
SUMMARY
The functional-justifier.tlon and ef fects on safety of reducing the number of SFRCS manual initiation switches from 10 to 4 is discussed in the safety evaluation for F0D 87-1107. The configuration of components on the new panel insert is within the scope of FCR 85-0109. Maintaining proper che.nnel ;
separation within the console and performing the seismic analysis of the reconfigure consoles is discussed in the safety evaluation for FCR 85-0109.
The effects on panel C5721 of the removal of the switches and viring and installation of the new plugs does not affect the structural integrity of the panel. These components are independent of other equipment located in the panel and, therefore, vill not affect the functioning of the other equipment located within the panel. The work associated with this FCR will not adversely affect the seismic qualification of the panel.
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.t As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
' The proposed' action vill not create the possibility for an accident or
, malfunction of a different type than any evaluated previously in the USAR, and does.not reduce any margin of safety as defined in the Technical Specifications..
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
cf FOR FCR 87-0092, REV. A, SUPP. 1 (SE 88-0107)
TITLE: 1 Modification of AFV/SFRCS Manual Initiation Switches. (Provide redundant .i L solenoid valves for the Main Feedvater Control Valves)
CHANGE:
Replace the existing Main Feedvater Control Valve Solenoid Valves SV-SP6A (SV-SP6B) with Redundant Solenoid Valves SV-SP6Al & SV-SP6A2 (SV-SP6B1 &
_SV-SP6B2)'.
l' REASON FOR CHANGE:
This modification is being performed.to reduce the possibility that the failure of a single solenoid or associated component (i.e.,.viring, power supply, etc.) vill result in the' loss of one train of main feedvater.
SAFETY EVALUATION
SUMMARY
This existing Main Feedvater Control Valve solenoid valves SV-SP6A (SV-SP6B) vill be replaced by redundant solenoid valves SV-SP6Al & SV-SP6A2 (SV-SP6B1 &
SV-SP6B2) which vill be pneumatically AND-gated and equipped with mechanical Eposition switches. In the current design, the logic is AND-gated i'n the SFRCS cabinets by. requiring two relays to de-energize to actuate each SV (SV-SP6A &
B). The new design vill maintain ~the AND-gated logic, but vill accomplish,it by providing two SV's in series, each of which requires de-energizing only'one relay to actuate. . The existing solenoid-valves.(SVOSP6A & 6B) and the nev
'SV's automatic controller positioning via the ICS syctem. When the existing SV de-energizes, its associated control-valve closes by venting the. pneumatic actuating circuit. This control philosophy rema5ns unchanged with the addition of the redundant SV's. SV-SP6Al (SV-SP6B1) vill de-energize to vent, but the. vent. path vill have another.SV added: SV-SP6A2 (SV-SP6B2) which also de-energizes to. vent, thus requiring both SV's to de-energize to vent the 4 pneumatic actuating circuit and close control valve SV-SP6A (SV-SP6By. Adding <
another SV in the vent path will allow one of the SV's at a time to be tested without interrupting main feedvater operation.
-The effect of adding two SV's in series on the vent path reduces the possibility of isolating one train of main feedvater-from the steam generator as a failure.of both SV's/ associated components vill nov be required to line up the pneumatic operating circuit for the Main 7eedvater Control Valve for closure. The other safety and operating features of the pneumatic operating circuit vill be maintained-(i.e., control valve fails as-is on loss of instrument air supply this maintaining backup accumulator air pressure for closure while minimizing feedvater system transients; and this modification has no effect on automatic control valve positioning via the ICS system).
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR>
The proposed action vill not create the possibility for an accident or malfunction'of a different type than any evaluated previously in the USAR, and-does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SUMMARY
- l. FOR FCR 87-0109 (SE 87-0286)
TITLE: I RPS High Pressure Setpoints and ARTS Threshold Power CHANGE:
Raise the Reactor Protection System (RPS) high pressure trip setpoint from 2300 psig to the original value of 2355 psig.
REASON FOR CHANGE:-
Following the Three Mile Island, Unit 2 (TMI-2) incident, to reduce challenges to the PORV, the NRC required that:
The PORV setpoint be raised to above the RPS high pressure trip setpoint.
At Davis-Besse, the PORV setpoint was raised to 2450 psig from 2255 psig.
A safety-grade automatic ARTS be operational. (NUREG-0737, Item II.K.2.10).
The RPS high pressure trip setpoint be reduced from 2355 psig to 2300 psig to further minimize challenges to the PORV.
These plant design changes have met the NRC guidelines for reducing challenges to the PORV. However, su', sequent operating experience has indicated a significant increase in t4e number of reactor trips. To reduce the number of 4 unplanned reactor trips ana reduce unnecessary challer. gas t.o sat'ety systems, '
the Babcock and Vilcot: Ovners Group (B&V0G) has previously proposed to:
Increase the turbine trip arming setpoint of ARIS to 45% of full power.
Raise the RPS high pressure trip setpoint from 2300 psig to its original value of 2355 psig. As a result of this proposed change, the minimum PORV setpoint allowed by the plant Technical Specifications has been raised to 2435 psig. The actual PORV setpoint remains at 2450 psig.
SAFETY EVALUATION
SUMMARY
Raising the high pressure setpoint to 2355 psig meets the NRC guideline for reducing challenges to the PORV following a high pressure trip. The PORV minimum setpoint which provides an 80 psi margin above the RPS trip setpoint is within the design basis of the Technical Specifications. Raising the ARTS-turbine trip arming setpoint to 45% vill reduce the number of reactor trips thereby decreasing challenges to safety systems. It is noted that the turbine trip arming restraint of ARTS to 45 percent of full power has not been implemented. Further, the overpressure transient for the control rod assembly withdrawal transient at startup is still bounded by the original FSAR analysis.
As summarized above, the proposed action vill not increase a the prob bili consequence of an accident or malfunction previously evaluatedtyinorthe USAR The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR does not reduce any margin of safety as defined in the Technical Specifications. , and Therefore, an unreviewed safety question does not exist.
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- .e SAFETY EVALUATION
SUMMARY
, FOR MOD 87-1011 (SE 88-0071)
TITLE:
Add-Local Flow Indication to the AFV Recirculation Flow Lines CHANGE:
LInstall a local flow instrument in each auxiliary feedvater pump recirculating
.. flow line.
REASON FOR CHANGE:
To meet the commitment of System Reviev & Test = Program (SRTP) item AF-NRR-006.
TheLlocal'flov-indicators vill allow the operators to check flovrates during pump runs and testing.
SAFETY-EVALUATION
SUMMARY
This modification vill enhance the safety aspects of the system by allowing a verification of recirculation flow. This does not affect any fire hazard analysis as there are no electrical components or combustible materials.used.
.The pipe hangers on:the EBD-14 (0) lines (which feed the HBD-137 (Non-0) lines) and the HBD-137 lines are being upgraded to meet the new seismic-requirements under this modification.
As summarized'above, the proposed action vill not increase the probability or consequcace of an accident or malfunction previously evaluated in the USAR.
The proposed action vill root create the possibility for an accident or malfunction of a different type than any evaluated previously;in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR MOD 87-1031 (SE 87-0396) ;
, TITLE:
Rod Stop Circuitry CHANGE -
i
. Add a rod stop circuit'in the Diamond Power Control Rod Drive System motor l control circuitry for.the safety, regulating, and Axial Power Shaping.(APSR). -!
rod groups.- 1 REASON FOR CHANGE:
To minimize the occurrence of uncommanded rod insertions, plant runbacks, and j reactor trips due to design limitations and component failures in the control j rod drive logic circuitry. 1 SAFETY EVALUATION
SUMMARY
The addition of a Rod Stop circuit to the motor' control circuitry of the Control Rod Drive system vill not affect the safety. function of any system. I The Rod Stop circuit vill have no effect upon the Nuclear Sarety-Related trip. ,
function of the CRDCS. The trip circuitry is physict.11y separated from the j system logic and motor control circuitry. The portion of the-CRDCS to be modified is not Nuclear Safety-Related. The Rod Stop circuit provides a manual actuation of the existing Direction Error relays in each rod group. A reactor trip can be initiated with the Rod Stop circuit actuated.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evalurated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different' type thsn any evaluated previously in the USAR, and does not reduce any margin of safety.as defined in the Technical. j Specifications.
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Therefore, an unreviewed safety question does not exist, t
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SUMMARY
TITLE:
l i
Upgrade Station and Instrument Air System CHANGE:
l Upgrade the existing Station and Instrument Air System by increasing the quality of the service air, increasing system capacity, and by providing greater system flexibility. This change includes retiring the old partial capacity emergency instrument air compressor, installation of a new centrifugal station air compressor, and reconfiguring existing SA compressors to serve as a full capacity emergency instrument air compressor.
REASON FOR CHANGE:
The system, in its present form, is undersized due to 2ncreases in demand.
Currently, a rented diesel-driven air compressor is providing the nccessary backup to the present system. Additionally, excessive moisture is being carried over to the station air header from the low point at the header drain valve.
SAFETY _ EVALUATION
SUMMARY
None of the engineered safety features depends on the supply of instrument air for its operatzon. All safety-related air-operated valves that arc required for safe shutdown are designed to assume a safe position by spring actuation and/or by a 0-listed accumulator system following a loss of instrument air.
These fail-safe functions are unaffected by this MOD.
The new Emergency Instrument Air Compressor (EIAC) and associated closed cooling loop vill have the capability to be fed from the Emergency Diesel Generators. Administrative controls vill ensure that the EDGs capacity is not exceeded if the EIAC is manually loaded.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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l SAFETY EVALUATION'
SUMMARY
FOR
TITLE:~
, ICS/NNI Auto-Select Svitch 1
CHANGE:
Incorporate automatic selection of valid input signals for the Integrated Control System (IC3).
REASON FOR CHANGE:
'The reason for this modification is to reduce the probability of a' plant transient caused by malfunction of a transmitter.
SAFETY EVALUATION
SUMMARY
l The Non-Nuclear Instrumentation (NNI) System is non-safety related,.but
' failure of NNI hardware >can initiate plant transients that result in challenges to the Reactor Protection System (RPS). Installation of the Smart Automatic Signal Selector (SASS) vill' reduce the occurrence of these challenges b.s' increasing the. reliability of NNI to ICS signals which~otherwise could challenge RPS if a failed signal were unmonitored.
.The electronic modules used for. SASS vill fail "as is" should. internal failures of SASS circuitry occur. With the. use of internalL jumper cards, SASS maintenance may be done bn line" without interrupting circuit. continuity to fany ICS inputs. Maintenance done "on line" could be accomplished without disabling SASS monitoring of more than four inputs. The' operator vill retain signal selection capability-of inputs even when SASS is disabled. ' This feature is such'that NNI system configuration and capabilities after SASS failure are.no different from system capabilities at.present. The impact on system operation of'a loss of individual signal vill be less than at present' when SASS functions normally, and no vorse *han present when SASS is not' operational due to' internal failt es or mairt nance in progress.
? As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safcty as defined in the Technical Specifications.
Therefore, an unreviewed safety question dc.es not exist.
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SAFETY EVALUATION
SUMMARY
FOR M0D.87-1093,'SUPP. 1 (SE 88-0183)
TITLE:
Control Rod Drive System Improvements CHANGE:
Install a Direction Error Detector module in the Diamond Power Control Rod Drive System logic circuitry for rod groups 5, 6, 7, 8, and the Auxiliary Group' Power: Supply.
REASON FOR CHANGE:
To minimize the occurrence of uncommanded rod insertions, plant runbucks, and
. reactor trips due to design. limitations and component-failures in the control-rod drive logic circuitry.
SAFETY. EVALUATION
SUMMARY
The addition of th'e CRD. Direction Error Detector vill not affect lthe safety function of any system. .The Direction Errot Detector vill be installed in the same slot as the previously utilized Diamond Power Direction Error Module.
This portion of the CRDCS is not Nuclear-Safety related. The Direction Error
~ Module will have no effect upon the trip function of the CRDCS. The trip circuitry'is physically separated from the logic circuitry.
L A. failure
. analysis of the (30) Direction Error module concluded that the. addition of the
~ Fault detector vill not result in an impairment of other CRDCS. functions,
.should the module fail.
As summarized above, the proposed actie vill'not increase the probability or consequence of an accident or malfunct..on previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does'not reduce any margin of safety as' defined in the Technical-Specifications.
Therefore, an unreviewed safety question dreo not exist.
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'[p SAFETY EVALUATION
SUMMARY
FOR H0D 87-1107, SUPP. 5 (SE 88-0100 REV. 1)
TITLE:
Decay Heat Removal Task Force and SRTP SFRCS Changes CHANG":
Enhance the SFRCS logics, input and output circuit interfaces to meet the scope of the modifications being made to improve the SFRCS.
REASON FOR THE CHANGE:
These changes are being made to reflect the recommendations of the Decay Heat Removal Task Force (DHRTF) ar.d the SFRCS Systems Review and Test Program (SRTP) Report.
SAFETY EVALUATION
SUMMARY
The fcllowing is a summary of each of the SFRCS enhancements being implemented under this modification. For the changes which have been evaluated in other FCR packages, the associated reference is presented.
- 1. 2-Out-of-2 Logic for Each Trip Input Parameter Modifying the SFRCS to trip on a 2-out-of-2 trip logic for each input parameter vill reduce the potential for spurious SFRCS trips due to testing or input sensor failures.
- 2. SFRCS Seal-In Feature,s The momentary seal-in circuit vill prevent an actuation channel from resetting following a trip for approximately 2 seconds. This vill assure all SFRCS actuated components respond to the trip signal.
- 3. SG Low Pressure Trip Logic The logic changes ensure appropriate AFV actions with both SG low pressure trip signals received sequentially. The modified logic maintains the SFRCS output trip signals, while both SG low pressure trip signals are present, ensuring appropriate actions by the AFV, based on the first low pressure signal received.
- 4. Shutdown Bypass Changes The elimination of the nuisance lights and alarms improves the man-machine interface.
The initial bypass circuits have no safety functions. Their purpose was to support the establishment of the shutdown bypass condition after an SFRCS lov steam line pressure trip condition. The capability to accomplish this function vill be maintained.
__ . _ , _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _m_._-. __m. _ - _ _ _ _ . - _ _ _ _ _ _ _ - _ _ . - _ _ - _ . _ _ _ _ . _ . _ . . _ _ _ _ _ - __ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ . , _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ ___..m_._.____-_____.._ _ _ _ _______.__;
- 5. Test Features This modification permits testing of the SFRCS function during shutdown in modes 4, 5 and 6 without the use of numerous jumpers, as previously required.
- 6. Deletion of' Half Trip Actuation This modification was made to assure that a closure signal to the listed valves was generated even if one of the SFRCS logic channels failed to trip. The circuits for the following valves have been rpgraded to full trip actuations: MS 100-1, MS 101-1, MS 1CS11A, MS ICS11B, MS 375, MS394, MS 603, MS 611.
This modification vill also reduce the potential for spurious closure of these valves during plant startup and shutdown operations when they may be open.
- 7. SFRCS Trip Signal Override The addition of inhibit circuits to SFRCS output trip signals for valves with "0 PEN" and "CLOSE" actuation signals, eliminates the potential of valve motor burn-out in an indefinable valve position. The failsafe signals on a loss of power to both complimentary channels vill automatically connect the steam supply from both steam generators to the affected AFPT and connect the auxiliary feedvater to the associated steam generator. This does not affect the accident analysis (i.e. main steamline break) in the USAR since this scenario only occurs following multiple power supply failures. These multiple failures are beyond the single failure assumptions of the accidents analyzed in the Safety Analysis Report.
- 8. Deletion of Eight Lov Steam Line Pressure Trip Switches The safety aspects of reducing the number of low steam generator pressure trip switches from 16 to B have been evaluated in the safety evaluction from FCR 87-133, the Technical Specification Change (Amm. 121) for this modification. The physical removal of these pressure switches will not impact the seismic analysis of the remaining switches.
- 9. Deletion of SG/Feedvater Differential Pressure 'nput Auxiliary Relays and (Sequence of Events) SOE Alarms The deletion of interfacing relays of the SG/feedvater differential pressure switch input circuits improves the reliability of these inputs.
With these modifications, the field contact vill be wired as a direct SFRC5 input, instead of through an auxiliary relay. The required functions of the differential pressure input signals are retained.
- 10. SFRCS/AFV Manual Initiation Switches The installation of the new SFRCS/AFV manual initiation switches has been evaluated in FCR 87-092 and FCR 87-130 (Tech. Spec. Change Amm. 124) for this modification.
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- 11. Blocking Capability of SFRCS Output Trip Signals l This Modification installs the circuitry and logic required to support I
installation of blocking capability for valves AF 3869 through AF 3872 and valves MS 106, MS 106A, MS 107 and MS 107A. The addition of this blocking !
capability is evaluated in FCR 87-0065 and FCR 87-0064. i
- 12. Control Circuit Changes The functions of the modified control circuits are identical to those of the existing control circuits. These modifications eliminate the chain of relays previously required to perform the functions. The new control circuits utilize the SFRCS output relays only.
- 13. Addition of a Second Solenoid Valve on Main FV Control Valves FUSP6A and l FWSP6B Valve response to an SFRCS trip will not be affected by this enhancement since both solenoid valves vill deenergize to close the valve as required.
The addition of second solenoid valve to FVSP6A and FWSP6B is evaluated in FCR 87-0092, Rev. A, Supp. 1.
- 14. Cabinets The new cabinets vill be color codad as the previous cabinets and are procured and installed to meet the seismic requirements.
- 15. Input Panel The enhanced SFRCS provides one input panel within each logic cabinet to facilitate periodic testing and maintenance of all monitored SFRCS inputs, and to revise the status of all SFRCS digital input signals through the see-thru vindow in the front door. The required testability afforded by the existing system vill be preserved and, in fact enhanced.
- 16. Output Panel The enhanced SFRCS provides one output panel vit'nin each logic cabinet to facilitate periodic testing and maintenance of all SFRCS trip outputs and to review their status through the see-thru vindow in the front door. The required testability afforded by the existing system vill be preserved and, in fact enhanced.
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- 17. Consolidation of SERCS Auxiliary Components The enhanced SFRCS eliminates from the control circuits all' auxiliary-relays interfacing the SFRCS relays with SFRCS actuated equipment in numerous relay cabinets at various locations.
_ SFRCS trip functions provided by the new control' circuits are identical-to those of the present.
system.
18.. Steam Generator Level Instrumentation
- The voltage values for the high and lov bistable setpoints as well as the change.in transmitter signal calibration range from 0-388" to 0-250" of indicated steam generator inventory is described and evaluated under FCR 85-0157.
Each signal monitor is provided with a qualified lE analog signal isolator. The outputs from these isolators are being.provided for future use, e.g., to. monitor the steam generators level signals during any mode of reactor operation with external test equipment. The isolated outputs from LT-SP988 and LT-SP9A8 are wired out to interface with the added barograph indicators at the Center Console. FCR 85-0096 provides detailed description and safety evaluation for this change.
- 19. AC Power Feeders for all Logic Channels The new power. sources are ultimately being supplied by the Emergency Diesel Generators.upon loss.of off-site power. The new power sources vill be. interrupted for up.to ten seconds until the Emergency Diesel Generators
- are on line. This load is not sequenced. During the 10 second interval, one logic channel in each actuation channel vill be deenergized. The other logic channels continue to receive power from the D.C. power supplies. The SFRCS vill respond as required to any trips initiators sensed in the remaining energized logic channels (i.e. loss of all four RCPs). This design and loss of offsite power scenario was evaluated in the SFRCS single failure analysis and found to satisfy single failure requirements.
The elimination of power feeders to external auxiliary' circuits will not affect the safety functions of the SFRCS since those circuits are relocated into the SFRCS cabinets.
- 20. Auctioneered Power Sources for Solenoids The enhanced SFRCS provides diode auctioneered power sources for all single acting solenoid circuits. The enhanced SFRCS circuitry will also provide positive indication on the output panel that the output relay contacts have changed state. Further a loss of a single power source vill not result in an actuation. Only the loss of both power feeders and/or trip of the two complimentary SFRCS logic channels vill result in an actuation.
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l These enhancements do not adversely offect the safety function of the Steam and Feedvater Line Rupture Control Systems (SFRCS), the Steam Generator Level l Instrumentation Cabinets (SGLIC), the Anticipatory Reactor Trip System (ARTS),
the Secondary Plant System including Main Feedvater and Auxiliary Feedvater Systems, or the Main Turbine T"ip System. The proposed actions vill not increase the probability or c-asequence of an accident or malfunction previously evaluated in the USAR. The proposed actions vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and do not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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t SAFETY EVALUATION
SUMMARY
FOR I MOD 87-1124 (SE 87-0395)
TITLE:
Control Rod Drive Cooling Vater System Modifications CHANGE:
1 Replace the existing CRDM cooling water lines with flexible stainless steel hoses and quick connect / disconnect couplings.
- 2. - Remove the restricting orifice (R0 2954) from the discharge line from the CRD cooling water booster pumps,
- 3. Reroute the minimum recirculation lines for the two CRD cooling water booster pumps.
REASON FOR CH!RGE:
- 1. This modification vill provide a simple and effective means of connecting / disconnecting the cooling water lines between the CRDM stators and cooling vater supply and return manifolds.
- 2. The restricting orifice is being removed to ensure that the required flov of cooling water is maintained with the quick connects in plcce.
- 3. The recirculation lines are being relocated to climinate back flushing of the inlet filters.
SAFETY EVALUATION
SUMMARY
The flexible hoses are being installed on a non-safety related portion of the system near the top of the CRDM's. The proposed configuration differs from the existing setup only in that the new hoses have quick disconnect couplings.
t These components are fully qualified for the service. The automatic shutoff characteristic of the couplings vill significantly reduce potential stator damage due to water spillage. No pipe whip or missile generator concerns are being created by this addition.
The restricting orifice being changed is on the line outside the functional 0 boundary (the MOV) but prior to the first anchor. The change from an orifice to a non-restricting spacer plate does not alter the line's seismic analysis.
The minimum recirculation lines' taps off the two booster pumps' discharge lines are being moved roughly four feet upstream of their present location.
No safety related equipment is located in the general vicinity of this piping.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed'actfon will not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USf3, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR RFM 87-1142 (SE 87-0136)
TITLE:
Generator Trip for Stator Ground Fault CHANGE:
Convert the existing "D-B Unit Generator Stator Ground" alarm function to a generator turbine trip function.
REASON FOR CHANGE:
The existing alarm function for generator stator ground fault condition is to be upgraded to a trip function to conform with standard industry practices of generator protection.
SAFETY EVALUATION
SUMMARY
This change vill improve the protection of the Generator Stator by isolating j the electrical ground fault in the stator from the turbine generator.
The existing generator stator ground fault protection is vired to alarm when a ground fault is detected in the generator stator vindings. A single ground fault, not cleared in a timely fashion, can cause significant damage.
Furthermore, a single ground fault increases the chances of a second ground on a different phase. An occurrence of a second ground fault on a different phase than from the first would cause the catastrophic failure of the generator.
The proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR. The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previnusly in the USAR, and does not reduce any margin i of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR MOD 87-1148 (SE 88-0069) i TITLE:
HPI and DH/LPI Flow Indicators CHANGE: i Replace a total of six (6) horizontal meters (GE Type 180) with linear scale Dixson bargraph style meters. The six indicators consist of four (4) High Pressure Injection (HPI) and two (2) Decay Heat (DH)/ Low Pressure Injection ,
(LPI) flow indicators located in the main control room on control panel C5716. l REASON FOR CHANGE:
The existing indicators have square root scales which make it difficult for operators to read at any condition less than 3/4 of full scale reading. Due to the need to perform flow balancing accurately as outlined in Attachment 6 of EP 1202.01, and to alleviate other Human Engineering Deficiencies (HED's),
the indicators shall be replaced with linear scale bargraph type electronic indicators.
SAFETY EVALUATION
SUMMARY
The new DH/LPI and HPI flow indicators are bargraph type electronic indicators made by the Dixson Corporation. The addition of these Dixson meters to the main control board and Foxboro signal processing cards to the Essential Metering Cabinets (ESM) vill enhance the operation of the flow transmitter indication loops. The Dixson meters provide better resolution and readability. Loss of input signal vill be detectable from the panel through the blinking of the bottom LED segmant. Power supply failures vill no longer generate mid-scale signal failures 1,r these loops.
As summarized above, the proposed action vill not increase the probability or consequerice of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
TITLE:
C.;ay Heat Removal System Valve Replacement CHANGE:
Replace the existing valve bodies, operators, solenoid valves, limit switches and associated equipment with new control valves HV-DH13A & B and HV-DH14A &
B.
REASON FOR CHANGE:
These valves are being replaced due to the frequent maintenance and repairs required for the existing valves and operator.
SAFETY EVALUATION
SUMMARY
The present configuration for HV-DH14A and HV-DH14B is as follows:' ;
- 1. One stem-mounted position switch provider position indication on a control room panel by a red (open) indicating lis;ht.
- 2. One stem-mounted position switch provides a computer alarm when the valve is not fully open.
- 3. The valve cannot be remotely closed without energicing the solenoid coil.
The power to the solenoid coil for the air control valve is monitored and a computer alare occurs when the coil is energized.
This modification elminated the diverse stem-mounted valve position limit switches, which provide computer alarms when either cooler outlet valve is not fully open. To provide equivalent administrative control of valve position, the local manual controls of the cooler outlet valves vill be locked so as to prevent either inadvertent or unauthorized local manipulation of the cooler outlet valves. By locking local manual valve controls, the cooler l outlet valves can only be manipulated remotely, which vould actuate the .i existing alarm relay. This provides redundancy to the stem-mounted positon monitor as recommended by AEC Regulatory Positoin 7.1.1.
The operation of the replacement DHR system control valves vill have no effect on plant safety, as the new equipment vill functionally be a one-for-one, like kind replacement of the existing equipment. The safety functions and fail-safe position of the new equipment will not be different than the existing equipment.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR,'and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, un unreaieved safety question does not' exist. ,
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SAFETY EVALUATION
SUMMARY
TITLE:
Modify ECCS Room Sump Pump Piping and Conduit CHANGE:
This modification. adds flanged connections to the six ECCS sump pumps discharge piping and shortens and relocates conduits for the pumps and'1evel switches.
REASON FOR CHANGE:
This modification vill facilitate maintenance of the pumps and associated sumps with a significant reduction in the man-hours required, while providing the capability-to remove any single pump without disabling the entire sump.
SAFETY EVALUATION
SUMMARY
These modifications vill'not affect the performance of the sump pumps or prevent them from performing their intended function. The addition of the flanges does not induce any additional pressure loss in the discharge piping and the minor reroute of the piping for pump P89-1A vill result in a smaller pressure loss thereby improving system performance.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
i FOR l MOD 87-1260 (SE 88-0061)
TITLE:
Add Condensate Removal Capability CHANGE:
Provide a steam trap and associated piping and valves to the main steam bypass line downstream of the MSR T33-2 supply line.
REASON FOR CHANGE:
Turbine bypass valve PV SP13A3 failed following the September 6, 1987 plant trip. An investigation was conducted to determine the root cause of the failure. This investigation concluded that overhanging of the valves' positioner cam caused the failure, but a contributing factor was water hammer caused by entrained water in the bypass steam. A design review of the main and bypass steam lines revealed a difference between the A and B sides. The second stage steam supply to MSR T33-2 is downstream of the steam traps for that train. Because of the reheater supply line's location, condensate can build up.
The addition of this steam trap will improve the reliability of turbine bypass valve PV-SP13A3 by removing condensate frca the steam supply line.
SAFETY EVALUATION
SUMMARY
This modification vill affect only the portion of the 12 inch bypass line downstream of MSR T33-2 second stage reheat supply line. This portion of the main steam system does not perform a safety function.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications. ]
Therefore, an unreviewed safety question does not exist.
t SAFETYfEVALUATION
SUMMARY
TITLE:
RPS Conduit Not Seismically-Supported CHANGE '
Conduit number 4-4642 presently routed through the Turbine Building is to be rerouted through the Auxiliary Building via Rooms 427, 422A, 428B, and 322.
- The. existing RPS cable number 4CRPSC04A'is to be relabeled as spare and coiled at both ends. A new RPS' cable number 4CRPSC04A is'to be pulled through the rerouted conduit number 4-46420A. This new RPS cable is to be terminated at both ends at the same termination points that the old RPS cable was terminated.
REASON FOR CHANGE:
In its present location, the conduit can not be seismically supported and is subject to high energy line break type accidents that could render the Channel 4 RPS circuit inoperable.
SAFETY EVALUATION
SUMMARY
Although this RPS trip is not required following a seismic or high energy line break, the route for the referenced conduit and cable is being revised to include rooms 427, 428B and 422A. These rooms were evaluated to determine the effects on safety due to the following hazards:
SEISMIC:
The referenced cable'is not required to be functional following a seismic event. However, the reference' conduit vill still be installed and seismically supported. Therefore, the referenced cable vill be able to withstand seismic events of the type that are in the original design basis of the plant and still remain functional.
FIRE:
Appendix R does not take credit for RPS to achieve and maintain safe shutdown.
In addition, a review has been conducted to ensure this cable is not included in listings for other Appendix R safe shutdown systems. Channel separation has been maintained.
RADIATION:
Room 427 is classified as a harsh environment due to high radiation levels.
However, these levels have been reviewed by the DED E0 department and were found to be acceptable. That is no degradation of the forty year life of this cable vill result from the radiation exposure experienced in room 427.
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~As-summarized above, the proposed action will not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR..
The proposed action vill not create the possibility for an accident or i malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an.unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
TITLE:i
' Remove the Internals of SV-329
' CHANGE:-
Remove the internals from'sving check' valve.SV-329 located in the' return piping from the Containment Air Coolers.
L REASON FOR CHANGE:
During normal' operationithis valve is subjected to lov fluid velocities which.
are not sufficient to maintain the valve fully open. This type of operation:
can lead to valve failure which could potentially cause a flow obstruction in' the~ service water return piping.
- SAFETY EVALUATION
SUMMARY
The function of SV-329 is to prevent backflow to the Containment Air Coolers.
'However this check valve is not needed for backflow protection due to'the presence of valves SV-1356 and SV-1366~vhich also prevent backflov'in this
.line. Service water check valve SV-329 does not' serve any safety function therefore' removal'of its internals will have no adverse effect on plant safety.
As. summarized above, the proposed action vill not increase the probability or:
. consequence.of an accident or malfunction previously evaluated'in the USAR.
-The proposed action vill not create the possibility for an accident'or ,
malfunction of a different type than any evaluated previously in the USAR, and
, , does not: reduce any margin of safety as. defined in the Technical Specifications.
. Therefore, an unreviewed safety question does not exist.
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L e SAFETY EVALUATION
SUMMARY
FOR j
MOD 87-1305 SUPP. l'(SE 88-0126)
. TITLES.
l-Modify Extraction Steam Control Valves CHANGE:
. Remove the'Limitorque motor operators and replace the present valve bonnets with a blind cover for the motor-operated non-return valves ES-264, ES 278, ES 370 and ES 377 :bi the Extraction Steam System.
. REASON FOR CHANGE:
This modification vill eliminate excessive forces on the rocker shaft and bushings which were created when the Limitorque operator was used to close the valves and_ isolate.the feedvater heaters.
SAFETY EVALUATION
SUMMARY
This modification vill increase station reliability by reducing the number of valve failures. The closure of the non-return valve to prevent vater induction into the turbines is not affected by removal of the motor operator.
The motor operator stem is not attached to the valve disc but only pushes on the disc to close the valve. The pneumatic air cylinder system which assists in, valve closure is totally independent of the motor operator.
As summarized.above, the proposed action vill not increase the probability-or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
FOR MOD 87-1317 (SE 88-0289) 1 TITLE:
345KV Generator Breaker Flashover Protection ~ !
I CHANGE:
Connect an instantaneous overcurrent relay PJC (50NT-2) in series with the current transformer circuit in the neutral of the main transformer. This relay is set to neutral current resulting from a 345KV breaker pole flashover.
REASON FOR CHANGE:
This modification enhances the existing breaker failure scheme for 345KV breakers 34560 and 34561 by providing protection against significant damage to the generator from breaker flashover which can occur prior to synchronizing or immediately after the unit is removed from service.
SAFETY EVALUATION
SUMMARY
The changes proposed by MOD 87-1317 vill not adversely affect the safety function of any safety related system. The modification increases the level of protection afforded to the unit generator against inadvertent energization by breaker head flashover.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The-proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
TITLE:
i Generator Inadvertent Energization i CHANGE:
)
l Add relays to detect generator accidental energization when the generator is off-line, when the generator is on the turning gear or when the machine is ;
rotating at sub-synchronous speed with the generator voltage less than a pre-determined value.
REASON FOR CHANGE:
To enhance the existing protective relaying for the generator by adding relays which vill function to protect the generator against accidental energirntion which could result in catastrophic damage to the machine.
SAFETY EVALUATION
SUMMARY
This modification vill not adversely affect the safety function of any safety related system. The modification increased the level of protection afforded to the unit generator by guarding against inadvertent energization. The relays added by this modification vill receive DC power through a separately fused circuit (tapped off of Circuit DBP01/DBN01). This coordinated fusing vill prevent faults within the inadvertent energization protective relays from causing a loss of DC power to other "B" generator and transformer relay protection schemes.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
b FOR' MOD 88-0049'(SE 88-0273),
H TITLE:
Orient Valve MS 5889A to a Vertical Position CHANGE:
Reroute the steam supply'line to Auxiliary Feedvater Pump Turbine (AFPT) 1-l' and reorient steam admission valve MS 5889A vith its valve stem in a vertical position.
REASON FOR CHANGE:
It has been determined that MS 5889A valve is leaking excessively. The leakage is apparently' caused by a slight misalignment of the plug and seating Lsurface. The misalignment is due to the weight of the pilot and main plug
' deflecting the stem which is horizontal.
cSAFETY EVALUATION
SUMMARY
There are no effects on safety as a result of this modification. Operation of' valve MS.5889A vill not change (e.g. stroke time, failure position).. The rerouted steam supply piping,-other piping, instrument lines, air lines, electrical conduits and hangers which vill be relocated and/or modified vill' not impact plantLsafety because they vill all be designed according to their
.present requirements (e.g. O, S/I) and their plumbing and viring schemes vill remain in their present configuration.
The postulated high energy line break (HELB) locations vill move with the relocated steam lines. Potential safety-related targets in the room were reviewed for the adverse effects of postulated HELBs. No additional protection is required as.a result of this MOD.
As summarized above, the proposed action vill not' increase the probability or consequence of'an accident or malfunction previously evaluated in the USAR.
The proposed' action vill not create the possibility for an accident or '
malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
t Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
TITLE:
SFAS Test Circuitry Viring CHANGE:
Revire the SFAS test circuitry to provide proper SFAS logic testing.
REASON FOR CHANGE:
The logic channels of the SFAS are made up of solid state components. Relays are used as terminating devices of the SFAS logic, as isolation devices for remote control pushbuttons, and as output signals to the station annunciator and computer.
The terminating relays of sensing and logic Channels 1 and 3, must both be de-energized to activate safety actuation Channel 1. Similarly, sensing and logic Channel 2 and 4 are de-energized to activate safety actuation Channel 2.
The terminating relays act on the actuation control devices such as motor controllers and solenoid valves.
Manual testing of the system logic is designed into the SFAS. Each two-of-four logic coincidence matrix of a system logic includes a local independent momentary pushbutton which, when operated, changes the matrix functioning to a one-out-of-four logic. This test with simultaneous presence of a bistable test trip on any channel vill de-energize the output relays of one channel of the protective action system being tested. Any combination of two or more bistable tr:ps of redundant channels associated with the same coincidence matrix logic vill cause a trip of the protective action system.
This modification vill change this testing viring. Presently the tested logic in each coincidence logic is initiated through components which are not sensing plant parameters. This change vill allow testing of the AND logic through components sensing two plant variables. This does not change the testing scheme, but improves reliability by testing actual components required to actuate the SFAS.
SAFETY EVALUATION
SUMMARY
This modification will result in greater reliability of the SFAS. This modification ensures testing of an active solid state logic responsible for an SFAS actuation. Presently tested logic involves AND gates which are not sensing two plant parameters. This improves reliability by permitting testing of an actual component required to operate to initiate a system actuation.
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This modification vill not result in any change in effects from environmental conditions. The components are identical and located in the same cabinets in a mild environment. Electrical noise vill have no greater impact than previously encountered from the bistable test switch. Noise would need to be present on two sources simultaneously to trip an AND gate and would have to occur in two coincidence logic matrices.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and 1 does not reduce any margin of safety as defined in the Technical .
Specifications. -
l Therefore, an unreviewed safety question does not exist. f l
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SAFETY EVALUATION
SUMMARY
j i FOR MOD 88-0234 (SE 88-0637)
TITLE:
Overpressure Protection for the Containment Air Coolers (CAC)
CHANGE:
l Add individual relief valves to each Containment Air Cooler (CAC) return line between the containment penetration and the service water isolation valve.
The installation vill utilize the pressure taps for the pressure indicators.
The pressure indicators vill be deleted.
REASON FOR CHANGE:
ASME Section III, 1971 Edition, Paragraph ND-7155 requires individual pressure-relief devices to be installed for the overpressure protection of components which are isolable from the normal system overpressure protection.
The CAC's are currently capable of being isolated from the service water system relief valves by closing their respective containment isolation valves (i.e. SV 1366 and SV 1356 for cooler E37-1). Therefore, the potential exists for overpressurization of the CAC.
SAFETY EVALUATION
SUMMARY
The new relief valves vill be added to the Class 2 section of the piping system. The relief valves are "0", seismic Category I, and protected against hazards such as pipe whip and jet impingement. The new reliefs are installed in a lov energy system therefore the addition of the reliefs does not create a whip or jet hazard for existing equipment.
The new relief valves do not perform an active function for isolation of containment, therefore the valves do not need to be added to the list of Technical Specification isolation valves.
CAC and SV systems vill not be adversely affected by the addition of the relief valves. The modification vill prevent overpressurization and potential i failure of the CAC's. {
I As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for ac. accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications. I Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
-TITLE:
Whip Restraints on the 10" Pressurizer Surge Piping CHANGE:
Modify the pipe whip restraints on the 10" pressurizer surge piping.
REASON FOR CHANGE:
It was discovered that the existing whip restraints did not meet the design as stated in the USAR. In conjunction with this issue an NRC concern on thermal stratified flow in the surge line was being evaluated. A review of these issues concluded that using more current pipe rupture analysis design criteria vould change the postulated pipe rupture locations s'ach that the existing pipe whip restraints vould not be required for pipe whip concerns.
No restraints are required for whip protection however, some of the restraints vill-be utilized to prevent the surge piping from falling if a break. vere to occur.
Piping thermal displacement in excess of the design basis movements is expected based upon the preliminary results of the thermal stratified flow study. The piping is being instrumented to record the thermal displacements at various~1ocations to verify these analytical predictions. .To minimize the
. potential for interference with the whip restraints under a stratified flow condition, several shims are being removed under this MOD.
SAFETY EVALUATION
SUMMARY
Theuseofthemorecurrentpiperupturecriteriavillhavenoadverse[ffects on safety.; Piping ruptures are now postulated at the points of highest probability for failure and if a break were to occur a review has been performed-concluding that no safety related equipment would be affected.
As summarized above, the proposed action vill not increase the probability or consequence of an accident or malfunction previously evaluated in the USAR.
The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not reduce any margin of safety as defined in the Technical Specifications.
Therefore, an unreviewed safety question does not exist.
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SAFETY EVALUATION
SUMMARY
TITLE:
Turbine Bypass Valve (TBV) Line Condensate Control CHANGE:
- This modification vill make the following changes to improve condensate collection and control
- 1. Add a bypass drain line in the piping lov point between each TBV and the associated downstream block valve.
- 2. Add a condensate collection drip leg in the low point of the 12-inch "B" l turbine bypass line (pipe elevation 598'-6"). The drip leg vill be provided with a condensate high level switch, automatic air operated blovdown valve, steam trap, and a manual bypass line.
- 3. Upgrade the collection drip leg for steam trap ST-143 on the 12-inch "A" turbine bypass line to include a manual bypass line, level switch, and automatic blevdovn valve.
- 4. Add a condensate level switch and an autop.7 tic blowdown valve to the trap varmup arrangement at the end of each TBV header.
REASON FOR CHANGE:
Improve the condensate removal capability downstream and upstream of the Turbine Bypass Valves to prevent damage to the valves due to the existence of water in the piping which acts as an inhibitor to steam flow.
SAFETY EVALUATION
SUMMARY
- 1. The bypass drain line has been analyzed for thermal stress, and its addition does not adversely affect the existing turbine bypass to the condenser line. The addition of this drain line vill improve the reliability of the turbine bypass valves.
- 2. The enhanced condensate removal capability vill not adversely affect operation of the Main Steam System. The proposed modifications are in the non-nuclear safety-related portion of the system. Spurious operation of the blowdown valve vould cause a loss of steam from the secondary system, but the cooldown rate would be bounded by existing postulated transients.
- 3. The new drain lines have been analyzed for thermal stress, and the additions do not adversely affect the existing 12-inch main steam piping.
l 4. The proposed condensate drain changes connect into the existing drains to l
the condensers. The drains have adequate capacity for the amount of l condensate to be removed, therefore, the modifications do not adversely l
affect the operation of the Main Turbine Seals and Drains System.
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l As summarized above, the proposed action vill not increase the probability or ;
consequence of an accident or malfunction presiously evaluated in the USAR. l
'The proposed action vill not create the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR, and does not' reduce any margin of safety as defined in the Technical l 1
Specifications.
Therefore, an unreviewed safety question does not exist.
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