NRC-91-0039, Safety Evaluation Summary Rept 1990
| ML20070S275 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 03/18/1991 |
| From: | DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20070S269 | List: |
| References | |
| CON-NRC-91-0039, CON-NRC-91-39 NUDOCS 9104020219 | |
| Download: ML20070S275 (106) | |
Text
{{#Wiki_filter:_ -. . _ ~ . . -- . .. . (2
.c' Enclosure to NRC-91-0039 l
FERMI 2 SAFETY EVALUATION
SUMMARY
REPORT 1990 Docket No. 50-341 License No. NPF-43 9104020219 910318 DR ADOCK 0500 3 1
.< ?
FERMI 2 SAFETY EVALUATION
SUMMARY
REPORT 1990 AS-BUILT NOTICES N
.s l
- MFETY EVALUATIONS AS-BUILT NOTICES page 1 SAFETY EVALUATION SO4MRY Safety Evaluation No: 89-0188 UFSAR Revision No. N/A Ref arence Docuesnt: ADN 10875-1 Section(s) , _
Table (s) Figure Change i 1 Yes (X1 No Title of Change: Ref ueling Area Inlet Isolation Valve and Reactor Building Exhaust Isolation Valve Drawing Changes SGMARY: This As-Duilt Notice revised the P & 10 Inc=atric Drawing. Functional Operating Sketches, and CECO data base to show - J proper QA Level, seismic category, ASME classification, and normal position / tailed position f or valves T4600F410 and T4600F407. The vendor drawings were also revised to show the correct PIS numbers for the associated solenoid valves and their mounting locations. T4600F410 is normally closed but automatically opens upon receipt of a low reactor water tevel, or high drywell, pressure signal and provides a ventilation flowpath from the refueting area to the Standby Gas Treatment System (SOTS). T4600F407 is normal,ly open and remains open during accident conditions to provide a suction path fnr SOTS operation. As this Built tiotico does not involve a physical o'.ent modification and the chantjes to the above drawings reflect the as-built plant configuration, these changes do not affect the accident analysis or the operation and function of any safety rol,ated equipment. w _ __.-______________________.________-__-___._________________m.__'_^ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - - - _ . _ . _ _ . _ _ _ _ _ _ _ _ _ ___.O
- .- - . ._ . - =-.-..-.- .- _..
^
'p s1 pb -
3 SAFETY EVALUATIONS AS-BUILT NOTICES Page 2 SAFETY EVAL.UATION SIAAMRY Safety Evaluation No 90-0014 UFSAR Revision No. 4 Re f erence Document : ADN 9889-1 Section(s) 5.2
~
Tabte(s) Figure Change [X3 Yes I 1 No 9 Title of Change Safety Relief Valve Modification SLAAMRY: In order to improve safety relief performance (specifically reduce setpoint drift) Detroit Edison:-implemented the Boil,ing - Water Reactor owners Group (DWROG) recommendations on this issue. These recommendations include:-
- 1) = on a test = basis, modifying 50% of the Saf ety Rollef Valves' (SRVs) pilot disc e asemblies with_a new precipitation hardened stainless steel materlat.
, PH13-BMo and, J) reduce labyrinth seat binding. In order to maintain the BWR00 recommendation that - 50% of the install.ed SRVs
- utilize'the PH13-9Mo disc meterlat, two- spare SRVs were modified prior to the RF01 outage and instatted during the outage (bringing the number of instatted SRVs containing PH13-8Mo pilot disc materiet to 7 of 15). To maintain this ratio.fseven of the spare pilot assemblies _ now contain the PH13-BMo disc materiet,-bringing the total number of. pilot assemblies containing the PH13-8Mo disc materiet to 14 of 31 total SRV pilot assemblies, spare and instatted. Att the pilot assembly tabyrinth' seats were inspected'and, as necessary, reworked.
For. each valve - the set -pressure -was adjusted - to within its proper value. rotested and racertified to be within accepted tolerances. Technical Specification 4.4.2.1.2 requires that half o'f the SRVs must be proven operable - at least once overy eighteen months = by performing a - set pressure test. The UFSAR is being revised - so that. It does not need a revision after
- each set- pressure test , perf ormed in accordance with Technical Specification 4.4.2.1.2. Also, Section 5.2.3.2 is being revised to identify. PH13-BMo as an acceptable material for exposure to reactor coolant,
.. - e e e --
4
.s
- b SAFETY EVALUATIONS AS-BUILT NOTICES Page 3 Safety Evaluation No. 90-0014 (continued):
Wdifying the pilot assemblies to bring the number of valves incorporating PH13-BMo disc material to 50% of the installed SRVs and inspecting and, as necessary, reworking, rotesting and recortifying those valves with labyrinth seat problems helps ensure the SRVs perform the design is Tction descelbed in Sections 5.2.2 and 6.3.2.2.2 of the UFSAR, and in t h' basis of Technical Specification 2.1.3 " Safety Limits for Reactor Coolant System Pressure", 3/4.4.2 " Safety /Retief Valves", and 3/4.6.2 " Reactor Coolant System Pressure". These changes are consistent with the functional operation of the NGSS system and are made to increase the operability and integrity of the Main Steam Safety Relief Valves and internal components. ______-_m_. -____.m
.s J
SAFETY EVALUATIONS AS-DUILT NOTICES Page 4 MFETY EVALUATION SLMMRY Safety Evaluation No: _90-0022 UFSAR Revision No. 4 Ref erence Doc 4snent t ADN 10372-1 Section(s) Table (s) 1.6-2 and 1.6-3 Figure Change IX) Yes [ ] No Title of Change Process Diagrams Change SUMMAY: It was identified that the RHR System prosess diagram contained in the UFSAR was not based on the latest revision or the correct drawing. Further. investigation showed similar problems with the process diagrams for Core Spray, HPCI, RCIC and RWCU. To correct this problem, the above re f erenced drawings have been converted from process diagrams to Detroit Edison drawings. The Edison drawings form the basis for the aforementioned UFSAR figures, and will be aester to maintain current in the future because updates to Edison drawings are routinely included in change document packages (e.g., E0P's). The changes made are based on previously approved documents or are editoriat in nature. The changes are required to update process figures contained in the UFSAR to reflect current plant design, and to provide a mechanism for updating the drawings as required in the future.
.o...... ..._ =........ ....... .............................................
4 SAFETY EVALUAT!DNS AS-BUILT NOTICES Page 5 SATETY L' VALUATION
- M RY Safety Evaluation No 90-0069 UFSAR Revision No. 4 Re f erence Doctament : ADN 11610-1 Section(e)
Table (e) Figure Change (X) Yes [ l No Title of Change: Locked Valve Program Changes SLD E RY: In response to Notice of Violation (60-341/80021-01). Detroit Edison agreed to re-review systems to determine if locked valves were required. A review of applicable systems was performed and Design Calculation 4959, Volumes i through 28, was issued to document the results. Design Calculation 4969 Volume 1 represents the Mechanical portion of the design cato, and Volumes 2 through 28 represent the 15C portion of the design cate. The design calculations identify several Mechenital and I&C valves that are required to be locked. Following issuance of Design Calculation 4959, changes were approved and imptomented to summarize thL results of the Design Calcutation and valves were subsequently locked in their in-service position. As a result of the review, changes were made to plant procedures and to system drawings to make them consistent. Locking manual valves in their in-service position provides additional assurance, through adherence to procedures, that operability of the associated system is enhanced by reducing the probability that a valve in that systen will be mispo itioned. Adding locking devices to valves wit \ secure the valves in their in-service position. This gives added assurance that systems and equipment important to safety are operable. The addition of locking devices does not affect the function of the v al,ve s (v other equipment) other than adding a physical barrier to help prevent mispositioning of valves. The weight of the Iceking devices is small compared to the weight of the valve and will have an insignificant effect on seismic consideratione.
-u.
1 s
;4%: +-
SAFETY EVALUATIONS AS-BUILT NOTICES Page 6' SAFETY EVAltJATION Sta4MRY Safety Evaluation No: 90-0083 UFSAR Revision No. 4-Reference Document: ABN 11364-1 Section(s) 3.8 Tabte(s) i Figure Change [ ] Yes (X1 No Title of Change: Control Center Pressure Boundary Concrete Structure Seismic and QA Level Upgrade StaAMRYa This-evaluation upgrades doors, masonry block watts, and penetration seats in. the control center. pressure boundary concrete structure which are not already
- designated as seismic category 1 and QA.levet 1_to seismic category I and QA levet 1. The changes include revising related design documents to upgrade the ;
above pressure. barriers to seismic category 1 and QA level 1 and revising UFSAR Section 3.8.4.1 to identify the control center pressure boundary masonry watts as seismic category-I to provide consistency with UFSAR Table 3.2-1. These changes do not involve a new mode of plant operation, a physical plant change, or reduce- the originat design safety. factors. The_ seismic qualification of the doors and masonry block walls f or ' upgrades to seismic category 1 is docurrented in design calculation DC 5129, Volume 1. The penetration seats. ara construct 6d similar'to other safety rotated penetrations and contain the same design details. QA Lovet 1 acceptance of these barriers is - based on the a s-bull,t configuration, materlat, construction, and stress levat. The as-bull,t. configuration--is in accordance with the design documents. The materiet is the standard commercial material. - f or this application.-is not subject to change of state or degradstion for the life of=the plant,-is located outside the harsh environment, and cannot be chaWnged by any accident. The .- construction and - quality- of the upgraded doors - ard seats is similar to the
. existing safety related doors and cenetration seats. The masonry watts were constructed under an approved procedure and are considered acceptable. The-stress = levels computed in DC- 5129A- are low compared to the anowable stresses-and,Ltherefore, provide.a sufficient safety. margin.
_ __.__________m____-_.-__m._
i.
.w e
SAFETY EVALUATIONS AS-BUILT NOTICES Page 7 MINOR ADN'S The following As-Built Notices (ADNs) resulted in UFSAR drawing or text changes. These changes were reviewed for potential safety consequences. Because the changes were minor and were made to reflect at-built plant conditions, a summary f or each was not prepared. The ABNs and their associated safety evaluations have been listed for reference. Safety Evaluation No.: 87-0308 Figure Ctange Imp 1ementation Document: ADN 8013-1.-3 Safety Evaluation No.: 89-0217 Figure Change Implementation Doctanent: ADN 11032-1 Safety Evaluation No.: 89-0218 Figure Change Imptomentation Document: ADN 11031-1 Safety Evaluation No.: 89-0235 Figure Change !~ Implementation Docunient: ADN 10852-1 Safety Evaluation No.: 90-0006 Figure Change Implementation Document: ADN 10835-1 Safety Eva18mtion No.: 90-0007 Figure Change Implementation Document: ADN 11056-1 Safety Evaluation No.: 90-0015 Figure Change Implementation Document: ADN 9103-1 Safety Evaluation No.: 90-0025 figure Change Implementation Document: ADN 11159-1 Safety Evaluation No.: 90-0027 Figure Change Implementation Document: ABN 11173-1 Safety Evaluation No.: p0-0055 Figure Change Implementation Doctmoent: ADN 11440-1 Safety Evaluation Ms.: 90-0056 Figuro Change Implementation Doctanent: ADN 9409-1 Safety Evaluation No.: 90-0082 Figure Change Implementation Doctanent: ADN 11404-1 r Safety Evaluation No.: 90-0084 Figure Change Implementation Document: ADN 1U543-1
E 4%'
=.
. . i' SAFETY EVALUATIONS AS-BUILT NOTICES' Page 8 MINOR ABN'S (Contirused)
Safety Evaluation No. 90-0094 Figure Change Implementation Document: ABN 11415-1 Safety Evaluation No.: 90-0111 Figure Change
- Implementation Document: ABN 4124-1 Safety Evaluation No.1 90-0113- Figure Ctange
_ Implementation Document: NsN 4148-1 Safety Evaluation No.: 90-0129 Figure Change '
- Implementation Docueent: ABN 10207-1
- Safety Evaluation No.: 90-0133 Figure Ctange
. Implementation Document: ABN 11419-1 Safety Evaluation No.: 91-0021 Figure Ctange Implementation Document ABN 3864-1 END OF ABN SECTION 1
~ r , -,s-. - . . . . . . - . - . - _. . _ - . . . -.- - - , . _.. . _ - ,.
FERMI 2 SAFETY EVALUATION
SUMMARY
REPORT 1990 ENGINEERING DESIGN PACKAGES
- - . . - . . ~ ~ - -- . .- - -.. -. .~ , _ - . . - - - - - - - - ---
2
;s~
SAFETY EVALUATIONS ENGINEERING DEB 10N PACKAGES Page 1 SAFETY EVALUATION StatMRY Safety Evaluation No 87-0053 UFSAR Revision No. A Reference Document: EDp 6498 Section(s) Table (s) Figure Change IX1 Yes [ ] No , Title of Change: Circulating and General, . Service Water Chemical Injection Systems S M Y: f
- . . .. i This modification provided for the instauntion of two separate . chemicat addition systems. The chemical inhibits carbonate scate formation and was added to the general service water (GSW) system and the circulating water (CW) system.' Components included storage tanks, variabla displacommnt pumps, l discharge piping with associated val,ve s , and flu connections with their !
associated vatves. Physical size of the tanks and actual location are the only
" differences between the systems. Monitoring systems were provided to evaluate the effectiveness of the chemical addition.
The - chemicat ' addition systems were - instaued to' assist' in preventing heat transfer surf ace fouting and improve - plant availability. In the event of a condenser leak, undesirable dissolved solids from the circulating water system wiu - enter the condensate. :The chemical,,.Powerline 3450, is expec ted - to - contribute less than IX . to 'the total dissolved solide in the circul,ating The - active - ingredient- in Powerline 3450. is an anionicaMy charged. water. copolymer which is removed - by anion. resin . in - the condensate polishers.. The polishers are expected to -prevent Powerline 3450 fr a reaching the--high-
- j. . temperature and pressure conditions in the reactor. However, iny resultant product that bypasses = the domineralizers and enters the reactor will have a -.
minimat or insignificant impact ' on reactor coolant chemistry, main'. steam radiation levels or.fuet cladding integrity. Powertine 3450 is an aqueous pH-5.3 sol,ution. It is. compatible with att gaskets and valve packing materlats.in the condensate, feedwater, and circulating water. system,. Powert,ine 3450 is not- , L. aggressive to admiratty brass or. cerbon steel,. In f act, Powerline 3450 is commonly used to enhance the performance of many corrosion inhibitors and will l assist' in reducing corrosion by minimizing' deposits and resul, tant under-deposit l; attack.
i4 l
,s SAFETY EVALUATIDNS ENQ1NEERING DESIGN PACKAGES Page 2 SAFETY EVALUATION StanMRf Safety Evaluation No: 87-0162 UFSAR Revision No. 4 Ref erence Document: EDP 7696 Dection(s)
Table (e) Figure Change t X) Yo.s ( ) No Title of Change High Pressure Coolant Injection (HPCI) Pump Discharge Check Valve Replacement StammRY: _
'This --modification ~ replaced the original Y.patteen piston type HCPI pump discharge check valve and its associated bypass hne with a swing type check valvo. This change was made to reduce . the potential for pressure transients experienced ; during perf ormance of the HPCI Reactor Vesset Injection Test.
24.202.01 on July 6, 1987.- Investigation of . this problem showed that the
-pressure transient was due, at least in-part, to the: failure of the discharge
; check valve tc etone fast enough. $1nce no mechanical problems were found.
Detroit Edison decided te replace. the- valve with a faster acting swing type
' check valve. The discharge check valve bypass piping was eliminated because
= the - replacement check valve's disc can be raised to allow.for filling and venting of the discharge line.
The' replacement. of the HPCI discharge check valve does not af f ect the design, operation, or rettability of the NPCI system.- The replacement valve meets or
-exceeds .QA., ASME code, seismic. environmentat qualification, and design
-temperature and pressure- requirements. To address potential . disc / arm separation _ problems that have - occurred at other = plants, the design of the disc /swingarm fastening arrangement was reviewed and discussed with the-replacement - valve manufacturer. -The manufacturer stated - that no1 separation problemst had been experienced with the fastening arrangement used in this check valve and ' that failures occurred with check valves used in continuous flow conditions rather',than the intermittent flow conditions thsc would be experienced in the HDCI system.
I
1 i SAFETY EVALUATIONS
. ENQ1NEERING.0ES10N PACKAGES j Page 3 SAFETY EVALUATION SLAAMAY q Safety Evaluation No: 87 0364 UFSAR Revision No. 4 Reference Document: EDP 1065 Section(s)
Table (s) j Figure Change (X1 Yes ( ) No t Title of Change: Oil Analysis Lab Modifications S M Y: .i
)
This modification so ~ id. an syswash station and provided a ventitation system for the oit' analysis Tab. The eyewash station is required due to the ' .emicate
- present in the . tab and per _ 0SHA - . A HEPA filter is instatted an the tab l
-exhaust of the ventitation system due-to1 the low radioactivity of the tube oit-samples.
This modification is an operational enhancement which will ' support more ! expeditious analysis - of tube oit. -There is no interface with -equipment important to safety. The quantities of cit in the tube oil samples are small, enough to have no effect on the combustible loading of the room. l (. l _, . . . _ , . . . , _ . . = . . . . , . _ . _ . . _ . . _ . . _ _ _ _ . . _ _ _ _ _ _ . _ _ _ ___..
+.. , - _ _ . . _ _ . _ _ _ - _ ..
t6. ' BAFETY EVALUATIDNS. ENGINEERING DES 10N PACKAGES Page 4 EAFETY EVALUATION SGAMMY Sefety Evaluation Not 87-0367 REV 1 UFSAR Revision No. 4 fief erence Document t EDP 8169 Section(s) ,_ Table (s) FigJre Change (X) Yes ( ) No Title of Changet FW Heaters EN. ES. CN. and 68 Draina Level Transmitter Replacement SLAAMRY: This modification replaces the original differential pressure type drains level
. transmitters on- f eedwater heaters BN. 69 6N, and 6S with c'isplacement type transmitters. As a result of changing out the transmitters, additional small bore piping was r equ i r e d . -.. This change. will enhance . plant rettability by reducing-spurious heater alarms and isolations.
This modification does not change- the operation or function of the affected foodwater heaters. The new transmitters are set up and calibrated so that they have the same input range, aero reference \svot, and proportional output as the
- originet . transmitters. The f eedwater heater loss and foodwater heater pipe break analyses remain unchanged.
J i sg 4 i SAFETY EVALUATIONS J ENGINCESING DESIGN PACKAGES i pag. s l l EAFETY EVALUATION SLhMARY Safety Evaluation Hos 88-0020 UFSAR Revision No. 4 q Reforence Document: EDP 8320 Sectton(s) ! l Table (s) Figure Change [X) Yes ( ) No
.1 Title of Changes Feedwater Hester Level Transmitter Changes l
MANARY: This modification replaced the non saf ety-related, non seismic, CA Level II, Class D tevet transmitters on the feedwater heaters from Foxboro modet 13A1.MK2-LLD-3TE differentist pressure type transmitters to Fisher model 2600T-2400 displacer type levet transmitters. This modification is designated non safety-related, non-seismic, OA Level II, Class D based on the existing heater drain system design requirements. The heater drain system is designed in accordance with ANSI D31.1 1973. Personnet radiation exposure during level transmitter calibration is increased by relocating the transmitters near the feedwater heaters as required for this type of transmitter. However, calibration of this type of transmitter is not complex and can be performed in a short period of time, personnet exposure can be minimized by performing the transmitter calibration during unit outages. This modification conforms to the existing design criteria for att interfacing equipment and areas. Replacing the existing dif f orential pressure type level transmitters with displacer type 1./et transmitters witt-eliminate transmitter reference tog errors, minimize the number of spurious alarm and control actions currently experienced on these feedwater heater levels, and enhance the prevention of turbine water induction. The replacement transmitters perform the same function es the original transmitters and are purchased to the same qual,i ty level.
__,_. ,._.._- _ _ . ~ . . ~ _ . . - . _ . . . . . _ - - _ _ _ . -
- 3. .-
4' l l SAFETY EVALUATIONS ! ENGINEERINQ DESIGN PACKAGES' Page 6. SAFETY EVALUATION SLAMARY
~
' ,af ety Evaluation No 88-0070 UFSAR Revision 'do. ,
4 Reference Document EDP 1809 Section(s) l Tabte(s) 9.3 1 Figure Change IX) Yes i 1 No t i Title of Change: Corn aion Product Sampler Panet
-i StANARY:
i This change provides f or instaue'. ion of a corrosion product sampler panet- f or condensato -and feedwater, insta untion of a new sample line for reactor foodwater, manifolding heater drain sample lines to-the f orward pumped drain
' sample line for remote monitoring, and instauntion of two-new rought. cooters away fekm high personnet-traffic areas.
The changes completed by this modification are s'trictly rol,ated to non-seismic, non-safety-related, ,OA Lovet _ II and III , components located in the turbine
. buil ding, The changes provide a better toot to monitor water chemistry and thus are~ considered an enhancement to the sampling systeta, l
I ': i p
. v.
SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES Page 7E SAFETY EVALUATION BLAAMRY Safety Evaluation No: 88-0224 UFSAR Revision No. 4 Reference Documents' EDP 1080 Section(s) 9.3 Table (s) 9.3-1 , Figure Change (X) Yes ( ) No Title of Change: Insta Mation of a Sample Drain Cou ection/ Recovery System SLBAMRY:
'This modification provided for the instaMation of a sample drain collection / recovery system for certain sample panels and sinks, re-routing of--
those drains to the sample drain collection / recovery system and re-activation
- of certain sample points in the turbine building.
' Addition of the sample drain couection/ recovery system enhances the radweste process. system by reducing the volumetric loads processed and reactivation of
- certain sample points provides more versatility to the plants = water quality .
analysis _ when . monitoring plant conditions and equipment performance. The modifications'made are strictly rotated-to non-seismic .non-safety related, QA levet II and III components located in the turbine buitdang. None of the , changes affacted the functional, or operating characteristics of the plant. 1
.Y t
,n-- n.-
SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES Page 8 SAFETY EVAL 1MTION SGNMY Safety Evaluation No: 88-0235 UFSAR Revision No. 4 Ref eronce Docueent EDP 7816 Sectlon(e) Table (s) 9.3-1 Figure Change [X1 Yes ( ) No Tit 1.e of Change: Turbine Building Sample System Upgrade SLS4AMY: Condensate dominera11rer sample drains and conductivity cetts were routed via floor drains to the radioactive waste system. Th19 modification rerouted these equipment drains to the turbine building sample drain collection tank. This modification provided an improved and expanded water quality monitoring capability for the process sampling system. This was an enhancement to the original system. The materials selected for this modification met or exceeded the original design requirements. The modification was strictly related to non-seismic, non-safety related. OA level 11 and 111 components located in the turbine building. The location of the major components and the necessary tube routing was selected with consideration for accessibil,ity, maintainabil,ity, and ALARA concerns.
-4
'I 4
SAFETY EVALUATIONS-EN0lNEERING DE310N PACKAGES Page 9 SAFETY EVALUATION SLM ERY Safety Evaluation No: 89-0028- UFSAR Revision No. 4 Ref erence Document s ,EDP 5299 Section(s) 9.3 1 l Tabte(s) 9.3-5 Figure Change (X) fes -( ) No Title of Change Reactor Buttoing Sub-basement Sump Modifications SlasMRY:
.This modification provided - f or the elimination of crossetto ' flood control valves and rerouting of: floor / equipment drains to' the tocal sump located in the corner rooms - of the . reactor ' building sub-basement, ' A1.so , the flood control valve tocated in-. sump D076 was replaced wi th a new motor operated - gate valve with an operator mounted above the aump cover.
~
The modifications reduce the'potentist for reactor building subebasement corner:
' room cross fl,ooding since some potentist pathways are eliminated, whi te others have pasrive barriers instaued in place of active devices. -The et*mination of
'the five flood controt valves reduced requirements for survetM ance testing and maintenance and additionaMy '. etiminated . au the associated ALARA concerns.
. Operating. _survel u ance, and maintenance improvements' associated with this modification result in economic benefits in the tong run.
m eme m eme eeee ee ewemeweeW 6e SO W emem e mm em em e m e mm e e ee m e m pee m e he m e tie mm em e e m e e me m e e 9 l
- l. -
r d
~m , . . - . , , , . . - ,
s, c
-{
(. f .
.9 -- SAFETY-EVALUATIONS
- ENQ1NEERING DESIGN PACKAGES-Page 10 SAFETY EVAlt1ATION SlaAMRY
+
4
, ; Safety Evaluation Not 89-0036 UFSAR Revision No. 4 Re f erence Doctament t .EDP 8484 Soction(s)-
Table (s) -
.i .
Figure Change (X) Yes [ ] No Titte of Change High Pressure Coolant Injection Room F1,ood Levet Instrumentation-- StaAMRY:
. This modification provided a high pressure coolant injection (HP01) room flood i level transmitter. ,controt_ room. annunciator, and recorder. Adding the instrumentation to monitor HPCI room fl,ood level, from tho' control room was >
completed. to- resolve c a= Detailed Contret Room Design Review . (4ROR) human engineering discrepancy.
-The' design change instituted by this modification is not saf ety-related . The ..
function-of the instaMation is to provide an additional means of monitoring HPCI room water level. Thare are no physical, connections between the new HPCI ;
- room. levet instrumentation and. any safety-related equipment. . Po.ential
-consequences-- of. room .ftooding wiu be reduced - since . earlier direct notification to the controt room is now - provided. - Att materlat instaued by this modification was - seismicaMy - mounted to ensure J that., other equipment
, important to- safety would not be adversely af f acted esia . result .of a seismic-ovent; n-Y'.
. . r b .. . ..m _w - ....- rw.- - - . ~ ~ ~ . . .
-. . . .- .~
s_
.=--
' SAFETY EVALUATIONS ,
ENGINEERING DESIGN PACKAGES-Page 11.- SAFETY EVALUATION SladdWRY Satety Evaluation No: 89-0041 UFSAR Revision No. 4 Ref er'snce - Document : EDP 10109 Section(s) 9.1 Table (s)- Figure Change (X1 Yes ( ) No
't Title of. Change Removat of Fuel Poot Cool.ing and Cleanup Vacuum Dreaker' Valves StammRY:
~
' This modification Peptaced the f ust pool cooling and cleanup (FPCCU) . system i return line ' vacuum breakers with open ended vent piping. The valves were replaced based on recommendations from the Technical Specification Improvement 1 Program.
- The replacement of the check val,ve assembl,1 s with piping enhances -the -ex'. sting piping system - by providing.. a vacuum breaker design that requires' no moving parte.. ThisJ vacuum breaker design is inherently- more reliable than the original check valves. The original purpose of the vacuum baeakers, to prevent-syphoning water from the f ust poot that wou'. reduce the water coverage over
. stored fuel less than 22 feet, la stitt maintained.
5
,_ f a
, ,q pm__ , ,- , - m . - . ,n,, - ,~v- e - -~r + - - - - - - - - - - - - - - - - -
h 3.. SAFETY EVALUATIONS. ENGINEERING DESIGN PACKAGES i Page - SAFETY EVALUATION AlaamRY Safety Evetuation No 89-0060 REV 1 UFSAR Revision No.- 4 Reference Document: EDP 8472 Section(s) 7.6 i Tabte(s) , Figure Change [X) Yes [ ] No _l Title of Change: Suppression Chamber Water Tsmperature Recorder Replacement ! SLAAMRY .
-This mor'.i f icat ian replaced a- suppression chamber water temperature recorder with -a reea' rder ' . having point, select and -averaging -capabilities. This modifiestion also changes torus water temperature annunciator window 8037 to an
~ amber window insreibed " Torus-Water Temp Troubto".
The . recorcer instal,ted by this modificatim added annunciation f or emergency
, operating procedure (EOp) entry . conditions that had not been annunciated .;
before. .,These alarm points w6re based on limiting conditions for operation i
'(LCOs) stat ed--in the technicaO specifications and EOP entry conditions stated in 7 plant .. procedures (Torus Water Temperature Control), The recorder- end.
annunciator points instened by. - this modification have no : cont rol - f unctions. g
. They simpty - annunciate high . torus water- temperature - conditions to atert - the l operators that an E0p entry condition has - been reached. This - provides - an -l improvement to--the design. !
i
)
.)
i 1
.. _ - _ - - = .l
7 t 4 SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES' Page 13 .
.i SATETY EVALUATION S M Y Safety Evaluation Not. 89-0066 UFSAR Revision No. 4 9efeverwe Document:- EDP 8576 - Sectlon(e) +
DER 88 074 i Tab.ats) Figure Change (X) Yes [ ] No
~
- Title of Changet Residual Heat Removal, (RHR) Suppression Poot Suction Valves, Shutdown Cooling . Suction Valves, and Test Return Valves Permissive Intertocks Instattation SLAAMRY:
This modification interlocks the RHR suppression -pool suction valves (F004A, B,C,D) and RHR test re turn ' valve s (F028A,B) with the'r respective shutdown - cooting suction vel,ve s =-- ( F006A,0.C. D) . - The purpose of - this interlock is to
. prevent inadvertent haining of the reactor vesset into the suppression pool, by
~
preventing the manuat electric opening of' either the suppression pool suction valve (F004) er the- test ' return valve when -their respective shutdown cooling suction vatva (F006) is.open. This modification is a result of an evaluation of-potentiat' reactor'.vosset drain paths conducted in response to Institute for Nucl. ear Power Operatfors (INPO) Significant Operett.ng Experience Report (SOER)
. 87-002.
t This. modification- does 7 not. change ~ the = design- basis, f unction, . or . modes .of
- operation of the subject valves or the RHR system. The F004 and F006 valves
= are not required to reposition!during or--after postulated = accidents. - The F028 valves are required to function after an accident and the ef f ects of-- a F006 1,imit switch f ail,ure were evaluated. 'The F006 val,ve-timit .at tenes - were
. upgraded to NURE0 0588, . Category 28 which qualifies them to . function under "y stulated accident ' environmental, and. seismic conditions. -
Therefore,- the probability - of ' f ailure ' of- the F028 permissive is no more likety than o t he r. - 4
-qualified-components and is acceptable, i
Y s-
tN ETY [ VALUATIONS [NG!b(( RING D[$10N PACKAGED Page 14 fMETY EVALUATION BLA4%RY Safety Evaluation Not 89-0071 UFfAR Hovision No. 4 Ref erence Doctsment s _EDP 8486 Section(s) __ Table (s) Figure Change (X) Yes ( ) No Title of Change Drywen and Torus Nitrogen Supply Valve Modification FABA MRY: This modi fication provided remote manual actuation capability to the drywett arid torus nitrogen supply valve. An air operated actuator was instau ed on the manual valve, Pushbutton contact and position indication were provided in the control room. This modification converted a normany closed manust valve to a remote 'oanual, fait-citsed, air-operated volve. It did not modify the systee piping configuration or operation. It improved the design by provic8 .ng remote operation capability from the main control room. The system af ter tod by this modification, the primary containment nitrogen and purge sys'em, is not required to operate during an emergency or fo' safe shutdown of tr. reactor and hence is not required to protect the health and safety of the public.
e. p ..- !L. SAFC1Y EVAWATIONS ENotNEERING DL810N PACKAGES Page 16 SAFETY EVAlt1ATION SLAAMMY Safety tvaluetto. No: 89-0077 UfSAR Revision No. 4 Ref erence Doctament: EDP 9979 Section(s) 3.68 3.98 4.33 4.63 5.23 6.53 EA.13 6.23 7.13 7.4 Table (s) 3.2-13 3.2-4 3.5-2: 3.9-18 3.9-73 E.2 13. 5.2-4 5.2-16 6.4-1 6.2-2, 6.2-16 6.2-16, 7.4 2 Figure Change (X1 Yes ( 1 No Title of Change Reactor Pressure Vesset Modifications , CAAamRY: This modification removed the portion of the reactor prensure vesbet (RPV) head spray tine betweer, the vesset head spray aorato end refueting pool floor. Blank flanges were instaMed at both the vesset head spray notate ftange and
-refueting pool floor penetrattoa. A set of blankect orifice flanges were
. Instened ' in the head spray piping upstroom of outboard isolation valve E1150F023 to prevent the drywen esction of the head spray line from fiMing with water if the drywen isotation valves leak. The orifice flange tape provide a high point vent for the Residual Heat Removat~(RHR) return header and a low point drain for the drywe M portion of head spray line.
Removat. of the head specy line does not af f ect RHR system rettability since it has no impact on the other operating modes of the system. No equipment ; important to saf ety is dep ndent upon the head sprey piping. Dy removing the flanged portion of head spray piping. RpV thermal duty win be decreased. i
4 SAFETY EVALUATION $ ENGINEERING DESIGN PACKAGCG Page 16 SAFETY EVALUATION SLAMARY Safety Evaluation No 89-0079 UFSAR Revision No. 4 ,, Reforonce Docusment: EDP 6546 REV 1 Section(s) , Tablets) Figure Change IX) Yes ! ) No Title of ChanDe: Main Steam Isolation Valve Pneumatic Controt Manifold Assembly Replacement SLMA8JtY: The main steam isolation valve (MSIV) pneumatic caa* rot manif old assemblies were replaced during the 1989 refueling outage with new qualified manifold assembtles. The environmental qualified life of the as senMies expired in December, 1989. This modification provided destDn. Installaticq and i0 sting instructions to replace the M31V manifold assemblies with now manifol.d assemblies. Tble modification ca1\ed for the replacement of the existing MSIV csntrot manif old assembly witn a new quotified manif old assembly. The change does not affeet the design basis. the manifold function, or the sequence or timine of MSIV operation. Therefore. att previous accident analysis results ,emain unchanged as a result of the new manifolds.
6 CAFETY EVALUATIONS ENGINEERING DESIGN PACKAGE 8 Page 17 Cart.TY EVAll1ATION PNRY Safety Evaluation No 89-0008 UFSAR flevisum No. 4 Re f erence Doctanent : I:Dp 78t_7 Sec ticm(s ) 9.3 Tablet e) 9.3-1 Figure Change IX) Yes ( ) No Title of Change Aeactor Duilding Plant Process Samp(ing System Modifications E N RY: The purpose of this modification is to improve and expand the water quality monitoring and control capabilities for the reactor plant process sample system (RPpSS). This has been accomplisbod by the addition of new monitoring devices and the replacement o' obsolete equifment. This modification did not af f ect any eqvynent which would adversely impact any previously evaluated acc> dent. The work performed and the materials specified met or exceeded the intent of the original plant requirements and are of a quality standard consistent with what has been used elsewhere in the plant. Interface with other plant systems, equipment and components has been considered in the design and the proposed modification does not alter the safety function of any system or equipment. The modification provides improved # avaltability and reliability and expands the water quality monitoring system for process sampling.
SAFETY EVAtVATIONS ENGINEERING DES!QN PACKAGES Page 10 qN MFLT( EVALUATION SLMRRY Cafety Evaluation No 89-0090 UFMR Revision No. 4 Re f erence Doctanent : FDP 9753 Section(s) Tabte(s) Figure Change (X) Yes [ ] No Title of Changes Seat 011 Coders and Stator Liquid Temperature Recorder Change StM MRY: Recorders monitoring the seal ott coolers temperature and stator liquid temperature respectively, were obsolete Bailey Modat 'SR" recorders. The servicing of theet devices was no longer feasible, since they were no \onger manufactured and spare parts were unavailable or had a 24 month toad time. This modifiestion replaced the Bailey recorders with Tracor Westronics Series 1100 (Modet 1110) recorders. These Tracor recorders are similar component replacements. The Tracor Westronics Series 1100 recorders have the same scale range as the previous recorders (0-200 F), accept the same 10-50 mA input signals, are color banded to the same canges, and, with an adapter plate, can be mounted in the same location as the recorders being replaced. The Tracor recorders are f unctionally and operationat\y similar to the Dailey recorders that were replaced. Furthermore, the Tracer recorders allow for a floating input signal, eliminating the need for the existing auto isolators. The types of oattunctions are the same for both the Tracor replacement recorders and tb ismiley t ecorders since they are att analog, single pen, strip chart recorders. The recorder installaticns are Seismic Category 11/1 as documented by Design Calculation 5022,
- l l
j 1 I i MFfh NAWATIONS I C 0 6.4H0 DCSIGN PACKAGES s av e 13 SAFETY EVALUATION Sta4MRY Safety Evaluation No: 89-0107 UFSAR Revision No. 4 Hef erenes Doctament t'OP 10400 Sectlon(e) Table (s) Figure Change IX1 Yes ( ) No Title of Change: Circulating Water Pump Suction Modifications Suh4MRY: The new design configuration, as described in this modification, was the instattation of an intake flow grid structure on the intake side of the five circulating water (CW) pumps. The purpose of the intake flow grid structure is to reduce or eliminate vortexing and extreme turbulence of the cooling water occurring near or in the pump sumps. Yhe grid is approximately 16' by 20' and sits in existing tracks for the stop gate. The grid structure consists of interconnected flat plates which form a pattern of l' by l' celle through which the pump intake water flows. The grid structure is OA \evet Ill, non-seismic, and is not required for the safe shutdown of the plant. The grid structure will cause a more uniform flow into the pump sumps. Therefore, the operation of the circulating pumps will be enhanced, helping to assure the c, ^1nuous rettable operation of the plant. This modification made o changes that will adversely impact any previously evaluated accidents. The work performed and the materlate speci fied met or exceeded the intent of the original plant requirements for CA Level 111 items. Interface with other plant systems, equipeent and components has been considered in the design and the proposed modification does not e\ter the safety function of any system or equipment.
. _ _ _ _ _ . . _ _ _ . .- . _ _ . - - _ _ _ _ _ _ _ _._ ~ _ - _. -__ _ . __ _ _ .
sc
- Y k
.. p f
SAFETY EVALUATIONS ENQ1NEERING DES 10N PACKAGES Page 20 I SAFETY EVALUATION BlatWtY
&afety Evaluation Not 99-0100 UFSAR Revision No. 4 Reference Documenti EDP 976T Section(s) u Tabte(s) _
Figure Change (X1 Yes ( ) No L Title of Change: Offges Preheater Outlet Temperature Recorder Change ; 1RAAWtY obsotete offges preheater outlet temperature Bailey recorders were reptoced , with Tracer Westronica Series 1100 recorders. .The auto isolators associated # with the Battey recorders were s'.iminated as they are not required f or proper operation of the Tracor recorders. t The Tracor recceders are f unctionetty and operationalty simitar to the Dailey recorders that were roptaced. The types of mettunctions are the same f or both the Tracor replacement recoeders and. the Battey recorders being roptered since they are aR anatog, single pen. strip chart recorders. The recorders are non-0 and their only function is to indicate and recoed offges preheater outtet temperature. They do not controt any other device or system. Functionat-requirements have not changed as a result of this modifacetion. f l
- . . ~ , . . , , . . . . .-a....,. ,,,_,__._._..n-- - - . . . . - _ . _ . _ - - , . _ . . - - , . - . . . . . . _ _ . . - .
*o fArrtY EVALUATIONS (NQ1NCERING DC810N PACKAOCS Page 21 n
SAFETY EVALUATION SLAAWty Safety Evaluation Not j 9-0112 UFSAR Revision No. 4 Reference Document EDP 5958 Sectionts) __ Tablets) Figure Change (X) Yes 1 1 No Title of Changes Main $ team Isolation Vatve Modifications SLAAWtY Thi s modification provided for roptacement of main steam isolation valves (MSIVs) internals and valve cover with parts modified to incorporate the valve manuf acturer's - ( Atwood & Morritt Co.) latest design enhancements which are i intended to improve tocal teskege rate testing (LLRT) performance. Tno replacement valve covers do not have a stem (packing) tenkoff connection. Therefore the associated stem teakoff piping is being removed for the MSIVs modified. The new MSIV parts are designed and fabricated by the original valve manufacturer to the originat design requirements but are modified to incorporate the latest design enhancements to improve LLRT results, mitigate rib wear and reduce any future potentist for stem fatture, and enhance operability and maintainability. The changes are improvements to existing design and do not affect the valves' function, operational performance or-qualification. Since the changes made~ have no affect on .the operationat characteristics of the valves.- but rather enhance their rettability both by improving packing perf ormance and eliminating potentist ce' tributing factors associated with the potential for stem failure, the potentist for valve fatture is reduced. b w., r. .., , -.
o 4 SAFCTY EVALUATIONS [NGINEERING DCSIGN PACF, AGES Page 22 tW ETY [ VALUATION BtMAARY Safety Evatustion No: 89-0116 UFLAR Revision No. 4 Referonce Docum.enti FDp 9794 Section(e) Table (s) _ Figure Change (X) Yes ( ) No Title of Change Deletion of the Off Oas System 20" Manifold Drains Cottoctor Tank " Low Lovet" Annunciatoe Input CAANARY: The new design deletes the Off Oas System 20" Mani'std Drains Cottoctor Tank Low Levoi ennunciator input which is one of the multiple inputs to Alarm Window 60t8. % 11guld in the 20" Manifold Drains Conector tank flashes away because of high ambient temperature and low pressure. Loss of liquid causes a constant Low Lovet Alarm in the Main Conttet Room. This liquid was originaMy intencoe to act as a sent to assure that any radioactive gases which may collect in the tank are contained within the Off Gas System and are not attowed to escape into other plant systema or the atmosphere. However, a liquid seat is not required in *.he 20" Manif old Drain Conector Tank bersune this (m.n is norma'Ay connected to the 20" Manif old which is maintained under the towest pressure in the off Oas System wl.ich is the towest pressure in the Fermi 2 Plant. The Off Oas System is non-nuclear safety-related. Att components affected by this modification are Non.0, Seismic !!/1 and Non-[0, The elimination of the Low Level Alarm and possible tcas of liquid seat in the 20" Manifold Cottoctior Tank witt not increas4 the potential for a release o' radioactive gases to the plant or atmosphere because any such gases in the Cottoctor Tank would be vented into the 20" Manifold due to the tower pressure within the Manifold. In this situation the 20" Manifold Co11ector Tank would be maintained at the vacuum tevet of the 20" Manif old or at a prensure between that of the 18* Manifold and the Condensate Receiver Tank, depending on the operating mode. This condition is rot a change to the conditions evaluated in the UfSAR. Additiontiny, this modification does not change the operation of the system, _ _ _ _ _ . . - - - . - - . _ _ _ _ . _ _ - - _ _ - - ______.__.-___-_______..____m_ _ - . _ m _-_.____.____---._____m_ __.._m._
.e
. J
]
SAFETY EVALUATIONS
- EN01HEERINO DESIGN PACKAGES ,
Page 23 l SAFETY EVALLMTION Sl4 MARY Safety Evaluation Not __89-0123 UlSAR Movisioe) No. 4 Ref arence Doctanont: EDP 9443 Section(e) Tablets) 1 Fleure Change 1X) Yes i 1 No Title of Change Rehester Seat Tank Level controtter Operation Improvement slaamRY: The purpose of this design change is to improve the stability of the reheater seat tank (RST) levet controner operation. This wee accomplished through-6 variety of piping changes, instrumentation changes, RST chLnges and miscousnoous changes. The changes are designed to restore the function o f . t he RST w'nich is to maintain watt seat between the reheater section of the moisture separator reheater (MSR) and the f eedwater heaters 6N & 68. Once the effective seat is maintained in the RST during operation. the possibility of reheater steam blowing into the feet. water heaters would be reduced. This can not be achieved by perf orming a single change in the RST. In this case, instattation of a baffte between the drains outlet and tevet sensing tap in the RST would reduce tevet fluctuation at the hvet sensing tap. Instauntion of a vortex breaker in the RST drain outlet would reduce the tevet turbulence in the RST by reducing vtsetex and steam entrainment in the drains cuttet flow, increasing. the RST vent piping diameter and valve stre to 2" wi n reduce the RST operating pressure facitatating continuous and steady stato draining of the east and west MSRs drains into the RST. Implementation of these three changes would improve the stability and performance of the RST tevet contro ner operation and thereby maintaina the water seat which is the original design basis for the RST.
-.ew a ,ww--,-- - ---~-w,-r---+,mw----,--=~---' -v-,-,-,-v
- re-t-r----w- -t e--m"---<rm'rmw ==--=*e-1e-*=- '-P-** ' ~ ' ' ' " " ~ "--""""-#' ' " " ' ' " ' ' * ~ -
9 SAFETY EVALUAT!DNS ENGINEERING DES 10N PACKAGES Page 24 SAFETY EVALUATION SlBM.RY Safety Evaluation No 89-0126 UFSAR Revision No. 4 Reforonce Document: EDP 10594 Section(e) Table (s) _ Figure Change IX1 Yes ( ) No Title of Changes HIAS System Receiver Tank Drein Line Supports SLheMJtYt The Division 11 Noninterruptible Control Air Supply System (NIAS) receiver tank drain pipe was found to be unsupported along its entire \ength (approximately 11 feet). Two supports shown on the piping isometric drawing are not instatted. The similar Division I drain pipe is supported at two tocations along its tength. A new design configuration, as described in EDP 10594, provides for the installation of two new pipe supports on the Division !! drain line. This will reduce the pipe stresses and deflections to a more acceptable working tevet, thereby increasing the design margin of safety and restoring the l aie to its reiginal intended condition. Please note that an analysis of the as-f ound condition of the unsupported drain piping concluded that the piping meets au licensing commitments and the NIAS system was operable. EDP 10594 also documents the design of the two existing supports on the drain line off of the N!AS Division I air receiver tank. Adding two pipe supports designed to seismic criteria to Division II drain line and documenting the design of the two existing supports on Division I drain line assures the piping stress tevels remain below the ASME 111 attowables. This assures the line is in agreement with the original UFSAR commitment to meet ASME III and the stress levels are made consistent with normal design practice. No existing safety related equipment function is being altered by this modification. l meeepeeeee&W#eSeeehneeemeeeeeeepppeee9mWW@#DSeeeeeeeem&9e&OeeeyeeeeemWheeeeemye
. _ . - _._ _ . _ _ _ . . . - - . = . _ _ _ _ . _.,
4 SAFE 1Y EVAL.UATIONS ENGINEERING DESIGN PACKAGES ' page 25 SAFETY EVALUATION SlatW4Y - Safety Evaluation No: 49-0132 Ur$AR Revision No. N/A Reference Docuumentt EDP 7819 Section(s) Tebte(s) Figure Change ( ) Yes (X) No Title of Changel Reactor Water Cleanup System Modifications IKeeWtY: The' reactor water cteanup (RWCU) system heat exchangers, both regenerative and non regenerative, were emperiencing -operational dif ficuttles due to periodic tenkage through tha ' f tange cover joints. The teakage was at tributed to the aging of the flat mete) diaphragm and .the diaphragm seat weld. This modification provides design.-instanation and testing instructions to replace the existing f1,at metal diaphragm by a thicker and targer diameter raised face metal diaphragm with.a new seat weld to prevent toskage through ; the ftange cover joint. l The modification w!M prevent future radioactive water teakage through tha heat r exckengers resulting in ' reduced plant contamination and radiation exposure.- i This wth also cut down' RWCU downtime, making' the system more available for plant support and reactor water oteanup. The change does not affect the design bases, functions, or- operation of the RWCU system. Au provicus accident analysis results remain unchanged as-a resutt-of-_-this modification.
?
+
- , or_~,.-,, 4._., , , - 4_,-,- , , . . . . - . . - - _ , , , , - . _ - . . . . - , , , . _ , _ . _ . - . . - - , . . . . - ._. - . . ~ . - - . . . . . - - . . . -
I SAFETY EVALUATIONS l EN0lNEERINQ DESIGN PACKAGES i Page 26 l I l SAFETY EVALUATION SLMMRY l i Safety Evaluation Not _ 89-0135 UFSAR Revision No. 4 1 Ref erence Document EDP 10622 Section(s) 1 Table (s) Figure Ctange (X) Yes ( 1 No Title of Change OPERATIONAL CONDITION 4, Shutdown Cooling Water Levet Lcw Atarm EMh84ARY: A Detailed Control Room Design Review Team was established to fulfitt the requirements of NUREO 0737 Bupplement 1 to identify and provide improvements in the control room that offer a high probability of f mproving plant saf ety by strengthening the man / machine interf ace. As a result of this review, several Human Engineering Discrepancies were identitled. EDP 10622 corrected one of these discrepancies by instatting an annunciator circuit which provides r:ontrol room starm during operationat condition 4 when reactor water level is 220 inches or less. Fermi 2 Technical Specification 3/4.4.9.2 requires reactor water levet to be maintained greater than or equal to 214 inches during operational condition 4 Previously there was no control room annunciator to alert the operators in avoiding the technical specification limit of 214 inches. The operator received the first warning cf tow reactor water levet at 193 inches. EDp 10622 instatts a new alarm input circuit and modifies the process computer and sequence of events recorder software to impecve plant safe operation by providing operators with advanced warning prior to reaching t he, Technicat Specification cold shutdown timit of 214 inches. This EDP does not modify any plant safety system since it affacts the balance of plant portion of the process computer software and annunciator circuitry only. The storm vonfiguration and annunciator to process computer interface used for this modification is similar to the existing annunciator inputs.
s.. SAFETY EVALUATIONS ENQ!NEERING DE810N PACKACEO Page 27 SArETY LVALLMTION StatMRY l Safety Evaluation Not 8g-0164 UFtAR Revision No. 4 Re f erence Domenent EDp 10714 Section(s) 6.2 Tabte(s) 6.2-2; 6.2 13; 6.2-16: 7.6-4 Figure Change 1X) Yes ( ) No Title of Changet Instattation of primary Containment Water Level Instrumentation SGSMRY A- Detailed Controt ' Room Design Review team was established to futfitt the requirements of NUREO 0737 to identify and provide improvements in the contret
~
room that of f er a high probability of improving plant saf ety by strengthening the man / machine interface. EDP 10714, in conjunction with previounty instau ed EDP 8483. corrected one of these discrepancies by adding instrumentation which wiu monitor drywen and torus pressure. This how instrumentation win aMow monitoring of primary containment water tevet up to elevation 650 f eet which was not previounty e.valtable . Primary containment water tevet monitoring capablu ty up to *he maximum floodable tevet of the containment is required by Fermi 2 Emergency Operating Procedures. The design of the sensing line is consistent with the requirements of General Design ' Criteria 54 and 56 and Regulatory Outdo 1.11 for instrument sensing lines penetrating primary reactor containment to ensure primary containment integrity and timit the potential off site dose below 10CFR100 requirement s. Solemic instattation of ennduit. cables recorder, transmitters, and instrument tubing ensures that the integrity of surrounding components or systems win not ; be impacted. The design of the dryweM pressure sensing line is consistent with the design of othee existing. Instrument tines preventing primary > l containment at Fermi 2 and f ans within the envelope of an instrument line pipe break accident scenario as described in UFEAR Section 15.6.2. Additionetty, the instaMation- of the drywen pressure sensing line does not connect to the I; reactor coolant pressure boundary and does not impact any component or system related to the safe shutdown of'the reactor. l l
SAFETY EVALUATIONS l (NQ1NEERING DES!QN PACKAGES , Page 28 SAFETY EVALUATION 9tateMtY Safety Evetuation No: 09-0170 UFSAR Revision No. 4 Referonce Document: EDP 10006 Soctionte) s Tebte(s) Figure Change (X) Yes ( ) No Title of Change Reactor Water Cleanup System Modification SLAAAARY: The reactor water cleanup (RWCU) bot tom head drain valve minimum - flow bypass valve was tenking past the seat and impacting the successful completion of the in service-' inspection tocat teak rate testing of _the inf> sed containment isolation valve. An engineering ovduation determined
- hat *tas valve was not required _ to support _ the operation of RWCU. Therefore this v.odification was prepared to remove the valve and cut and cap the associata s bypass lines as close as possible to the bottom head drain tino.
The RWCU bottom head drain minimum flow bypass vatve is inside containment and
- is bounded by the existing pipe break analysis within containment. Elimination of the minimum flow Une and replacement with pipe caps reduces the fatture modes f or the , system and subsequent probability of occurrence of an. accident.
Thus, au previous accident analysis results remain unchanged. Removat of the valve and bypass line wi n not affact the design bases. functions, or operation of the RWCU system. }
, r , , , . . ~ , . - t , . - _ . , . 1----,-e-,,_,.--.. ,,,,.__--,--,_...~m._,.,-_,,.,.,,_,,.-. - - . _ - _ , . _ . . _ _ . . , - _ , . . . , _ .~m__
SArtTY EVALUATIONS ENQ1N[ERING DESIGN PACKAG[$ Page 29 SAFETY EVALA1AT10N SLMAMY Safety Evaluation Not 89-0175 UFSAR Revision No. N/A Reference Document __EDP 7964 Section(s) Tab \e(s) Figure Change ( 1 Yss IX) No Title of Changet High Pressure Coo \ent Injection Valve Modifications SLA M M Y. The motor for the actuator of a high pressure coolant injection (HPCI) turbine injection valve motor operated valve (MOV) was revised from 100 foot pound starting torque (7.22 HP) to 160 foot pounds starting torque (10.83 HP). Also changed at the same time were the fuses (FRS-R-50 chanDod to F fSS- R- 80 ) , the fuse etips (60 amp to 100 amp clips) and the overload heater (changed to 030T53A). There was no impact on battery loading or sizing created by the change in motor size because this valve was origina\ty designed to operate at 150 f t. \b. The required testing and supporting changes (fuses, overload heater, limiter plate) were peat of the design change. The operability of the HPCI system was enhanced by providing more assurance of operability of the valvel thereby lessening the consequences of an accident. ~~
t# $TY tvALUATIDhe t".01f1EEAING DESIGN PACKAGES Pqr 30 fWCTY EVALUATION Sta4MJtY 3afety Evaluation No 89-Ot77 UFSAR Revision No. 4 Ref erence Doctament : EDP 10746 Section(s)
~
Tabte(s) Figure Change IX) Yes ( ) No Title of Change: Non-Safety Related DC System Ground Detection System Circuit Change SLM MRY: EDP 10746 changed the ground detection tight assemblies, such that an existing ground detection push button must be depressed to enable the ground detection tight system. This change remutts in the removat of the permanently instatted resistive ground path from the DC sys tem. This ground path was a concern of NRC Inf ormation Notice 88-86 Supplement 1 ' Operating with Multiple Ground in Direct Current Systeme". This notice dascribed the previounty unidentified potentist fatture mode, where a solenoid would fait to the energized state because of a circuit existing through the notonoid and the resistance of the ground detection lights. Thus, EDP 10746 provides ground detection capability and removes the instatted ground. The new design is en enhancement because it makes a component fatture (DC solenold, relay, etc.) to the energized state less ttkety. The ground detection light system operates as before. The only dif f orence is that the previous permanent 1.y maintained ground path witt only exist where the push button is depressed.
l . LVETY EVALUAT!DNS ENGINEERING DESIGN PACKAGES Page 31 SMETY EVALUATION DlMARY Safety Evakuntion No 89-0100 _ UFSAR Revision No. 4 Referonce Doctenent: EDP 1708 Sectionta)
~
Tif te(s) Figure Change IX) Yes ( ) ho Title of Change Mimic panet Replacement SLMMRY: The reactor iso \ation valve mimic display was replaced with a completely new display penet and controt cabinet due to the unrollability of the previous equipment. The dynamic flow portion of the reactor isolation valve mimio display was intended to be a quick reference system for the cperators. Alt valve position and system flow information provided on the dynamic flow portion of the mimic is availab\o elsewhere in the control room. Deletion of the dynnwie mimic does not modify or affect any plant system oe equipment other thnn the mimic panel itself, and it eliminates a potentially misleading source of information thereby enhancing the quality of information provided in the control room.
i.
. i .. l l
4 SMETY EVALMT!ONS ENGIN(( RING DECION 7% 4 AGE $ I Page 32 I l l IAF ETY (%llMTION tRERY j Safety Evaluation Nol 89-0164 UFFAR Revision No. 4 j Reference Document: EDP 9338 Deciion(e) Tabte(s) Figure Change (X) Yes ( ) No } Title of Change: Drywell Sump Modifications tRA4ANY: The drywell equipment drain sump and the drywell floor drein sump di scht rge flow instrument toops were modified to eliminate falso flow indications occurring at the drywell sumps' flow recorder and flow integrators. The drywett equipment drain sump pump control logic was also modified to eliminate actuation of a controt room annunciator window, whenever the sump water was being recirculated through a cooler. This modification does not change the original intent or function of the drywett sumps flow instrumentation. The required indication, recording and integration functions are maintained. The purpose of t he modification in to reduce, if not eliminate, the noise in the loop. Removat of the noise will eliminate the falso flow indication on the recorder and integrators, This modifiqation will not impact the safety function of any safety related system previously evaluated in the UFLAR. nor does it impact the operation of any othea plant systems or components. This change represents an improvement in performance and reliability. 1
e
&AFETY CVALUATIONS i (NGINCCRING D(S!QN PACKA0(S Page 33 I BArETY EVAL.tlATION 9ttemRY Sefety Evaluetlon No _89-0190 UFSAR Revision No. 4 _
1 , Reference Documentt , _ EDp 10666 Section(s) Table (s) j 4 Figure Chenee IX) Yes ( .1 No Titte of Chenpe ' Standby Qas Treatment System Carbon Dioxide (9078 CO,) i Tanks Lovet Indicators Modification
SUMMARY
)
l The originat levet indicators on -the BOTS CO . Tanks were float type mechani sms and were are an integrat part of fhe CO tank. With t ?.i s 2 conf.tgura t ioti the .tevet . indicators cannot be isolated f rom - t he tank and theref ore cannot be periodicatty calibrated without deatning t he -- t ank . The current des 10n for CO tank 'tevet . indication . by the storage tank's l manufacturer uses a differ,entist pressure indicator. A conversion is avattable 9 and recommended by the manuf acturae when periodic catibration of t he tevet, indicator is required. .EDP 10666 removed the original indicator and instatted differentia 1' pressure indicators on both 60T8 CO, Tanks (Division ! & !!), i Tl a new indicators meet t he tank manufacturer's requirements for fire. protection service and enhance the CO system operability- by providing ; reasonable assurance of . rsading accurate , tank t iquid - invent ory. The fatture ' 4 ef f ects of the new equipment instatted on . these CO tanks are bounded by the 2 existing arrangement, and the new tevet indicators do not adversely impact the operation of the CO system. 2
,%.-=,...c..p,y,y p...e,..... . , , - , . ..y..,-r.y ,3,y,,..,,,y.,m.y, , . ,,..,.wy ,-,,--.,_..-,,,,.,.-.y -
_ , - , . . , .,m . ...
4 SAFETY EVALUATIDNS ENGINEERING DESIGN PACKAGES i Pope 34 FNETY EVALUATION SLMY Safety Evaluation Not _ 89-9212 UFSAR Revision No. 4 Ref orence Doctanent : EDP 10962 Sectionte)
~
Tahte(s) Figure ChanDe (X) Yes ( 1 No iltte af Change Scram Discharge Volume Header Modifications SLMY: This modification provided on additional exhaust air flow path in parattet with the air regulator f or the scram discharge volume header drain valve actuator. In addition, applicable design drawings were revised to reflect the current plant confiDuration. This modification cetted for an addit;onal flow path for the exhausting of the scram discharge header drain valve actuator controt air only and does not a change the design basis, the valve function, or the required sequence / timing of the controt rod drive (CRD) hydraulic system operation. The risk of failure f or the check valve is no greater than the existing air regulator. Also the valve faits closed on a loss of power or a loss of air which is the only safety related function of the valve (pressure boundary after scram). Therefore, att previous accident analysis results remain unchanged as a result of the new modifications. I i t i l l
_ . _ _ _ m.__-.m ._m - - _ _ _ _ _ _ _ . _ . _ . _ . _ _ . _ . _ _ _ _ _ a. SAFETY EVALUATIONS ENQINEERING DESIGN PACKAGES Page 36 l l SAFETY EVALUATION StaaMMY ; Safety Evaluation No: 89-0214 REV 2 UFSAR Revision No. 4 , Refonance Docueent: EDP 10966 Section(e) , Table (s) - Figure Change IX1 Yes ( ) No Title of changet Modification of ConteM Air Supply to SCRAM Discharge Volume Header Vent and Drain valves i StaaMRY: EDP 10g66 was written to modif y the scram discharge votume (SOV) headee vent and ~ drain valves' (C11F160 and C11F181) common controt air supply. SpecificaMy, this modification replaces dual sotonoid valve C11F182 with two (2) eing,te colt solenoid valves, C11F182A and B which perform the same function as C11F182.= Both the C11F180 and C11F181 valves are controued via a single common duet 3-way solenoid valve (C11F182) which either supplies controt air to the valves common conte _ot air line et 76 psig when energized or purges the same tine to atmosphere when both cos ts are de-energized. The reason for this morlification was that the original solenoid valves (C11F182) port diameter ov9rty restricted air bleed off such that valve closure times did not meet Technical Specification criteria. This modification eliminates the design requirement to . close the SDV outboard valves 6 seconds after the inboard vatves.
- The new design uses two three-way, single coil, solenoid valves with larger port diameters connected in series in place of the existing dust coit ' solenoid valve. - The replacement solenoids utilite- the same colt voltage supplies and
- topic.
I The use of two single coit valves is f unctionauy equival, ant to the existing one dual colt valve. The existing dual colt valve requires that both colts be de-energized. (i.e., that both the Reactor protection System (RPS) Channet A & B scram signate be generated) in order to vont the air header supplying
- j. C1100F180 and C1100F181. Since the two roptacetrant single colt valves are l: mechanicany. connected in series. with one being actuated by RPS channeU A and-l the other by.RPS channet D. the 2 valve configuration wiu perf orm exactly as l the 1 valve configuration did. No adverse impact is created on the power supplies or logics since the replacement colts have the same electrical characteristics (i.e., as the two colts in the existing dual colt val,ve).
__ . _ . - . _ _ _ _ _ _ . _ _ _ _. ___ _ _ . - _ . . , -_. , _ _ _ _ _ . . ..~,_ . . _ . _i
*s
.u SAFETY EVALUATIONS ENQlNEERING DESIGN PACKAGES Pope 36 Sefety Cvetuation No. 89-0214 REV.2 (contirwed): The 6 second closing dif f orence between the SDV inboard anti outbosed valves was a design requirement to prevent water hammer. Dated , the conc \ution of Design Calculation 6120, the $ second time delay c\osure dif f erentist between the outboard and the inboard valve is hot required.
...................--......-------------------~~------------*------*------*----
I I l
ag' 4
. SAFETY EVALUATIONS ENQ1NEERING DES!QN PACKAGES Page 37 SAFETY EVAlthLT10N SLt4MRY Safety Evaluation No 89-0221 OrtAR Revision No. 4 Reference b _-- ,t EDo et/2 Bection(s)
Tab \e(s) Figure Change (X1 Yes ( ) No Title of Ohange Deletion of the Air Accumu\ator and Associated Equipment for. Reactor Feed pump Win. Flow Valves SLANARY: EDP 8372 rep \ aces the reactor feed pump (RFP) min-flow valves. This EDP deleted some control itemn which are no longee required with the new va\ves but are shown on UFSAR Figure 10.4-10. The deteted items are an air accumu\ator tank, a pneumatic so\encid vetve and several other pieces of equipment. Deco is del,eting this equipment es a precaution to peevent a potentia \ malf unction that ceutd f orce the min-flow valve open at a time when it should be closed and this cou\d reduce reactor water levet and possib\y initiate a Level 3 scram. None of the transients / accidents in Chapter 16 of the UFSAR take any credit f or the - RFp min-flow va\ves. The equipment being de\etod . is redundant to the replacement . valves f ait open spring. De\eting the extraneous equipment does not create the possibility of any new type of equipment malfunction.
4 9 SAFETY EVAWATIONS ENQ1NEERING DESIGN PACKAGES Page 36 SAFETY EVALLMT10si 9tteWtY Safety Evaluation Not 90-0012 UFSAR Revision No. 4 q Reference Document: EDp 8057 Section(s) 11.4 Tabte(s) Figure Change IX1 Yes ( ) No Title sf Change Primary Containment Radiation Monitoring Subsystem SlaAMRY:- EDP 8067 makes permanent the changes made in Temporary Modification. (TM)
- 87 300, removes the abandoned Hoffman enclosure which formerly housed the moving particulate assembly. removes att disconnected wires and removes att unused equipment which was spared per EDp 1786 frem the electronic and conteot anotosure. Hard ' piping instatted por TM = 87 300 is removed by EDP 8067 and replaced with new rerouted piping and supports. The particulate / lodine fitter 1s retocated to a more access.ble tocation to minimize the time required to change the fitter element.
Removat of the moving particulate f uter unit and replacing it with the appropriate pipe' etiminates a tenkage problem. The raoving partisutete fitter unit and. associated instaued spara equipment are no longer in use and served no purpose in support of plant operation. Instauntion of the new piping, .the j
-relocation of the particulate / lodine filter to a more accessible - tocation and the removat of non-f unctional spared components win not af f ect the originat Primary Containment Monitoring System design intent and perf ormance since no interaction between systems is offacted. The retocation of .the iodine fitter wi n reduce the ~ time requsred to change the fitter paper thus reducing radiation exposure to involved personnet.
. - - r - -- -
4 SAFETY EVALUATIONS ENQlNEERING DES!QN PACKAGES Page 3g SAFETY EVALUATION 9tMWtY _ Safety Evaluation Not 9C-0034 UFSAR Revision No. 4 Ref orewe Document EDP 11199 Soction(s) Tablets) , Figure Change [X):Yes [ ] No Title of Chenees Installation of Non Interruptible Air - Supply System's Accumulator Isolation Vatve Test Connections 9tM WtY: In _ order ' to _ completely test the Non-Interruptib\e Air Supp\y (NIAS) System, three test connection essemb\1on were instatted in accordance with EDP 11199, The new test connection assembtles wilt meet GA Levet-! Group C. ASME 111 Class t,' Seismic Category 1 requirements with the exception of the threaded pipe and pipe caps ~ downstream of the shutoff valve which are OA Level *NQ'. The
;. modification is in compliance with att the cetginal system design requirements, Omemeese emeese WOOme sewatse Opme emeeeeee e ese Oeepeem me e ewee e e ese e we e en e e ogeeemstegpe espee W ' ' ' . _ _ . . _ _ . -
.. j EAFETY EVALUAT10NS ENGINEERING DES!ON PACKAGES !
Page 40 SAFETY EVAltlATION staAWtY Safety Evaluation No 90-0046 UFSAR Revision No. 4 _ Refarence Document: ED4 7703 Section(e) Tabte(s) J Figure Change (X) Yes ( ) No ! Title of Change Addition of a Body Food Systwm to the Condensate Polisher Domineralisse System 3 1RaewtY This modification adds a body food system to the condensate polishing domineralizers. This consists of adding a mixing tank fitt tine, mixing tank drain piping, and individust vesset tnfluent feed tubing. . Body feeding consists of motoring amatt amounts of domineratiser resin to the fitter columns (septs) during domineratiser operation. Body f ooding compensates f or precost imperf ections, promotes depth filtration, and reduces the dif f erentist pressure buitdup rate. - As a resutt. fitter perf ormance le improved and domineratitor i precoats are reduced. The addition ~ of body feeding is within the design resin toeding of the domineratgrees because the inittst precost tending is tower (0.16 W/f t , vs. 0.208/ft. ).- Trw loss of the body f eed system witt not impact any previously evaluated accident anatyees ce af fect saf ety related equipment. If intertocks f ait and resin is continuounty f ed to the dominoratisers. the vessets will be removed from service due to high differentist pressure. Any pipe breaks in the body f eed system witt 'not impa6t the reactor bul'. ding teskage analysis The addition or _ tons of the body food system d6es not change the t echnicet specification chemistry limits. Y
SAFETY EVALUATIONS ENQ1NEERING DEST } PACKAGES Page 41 , SAFETY EVAlt1ATION (MANARY Befety Evaluation No 90 0061 UFSAR Revision No. _ N/A Reference Documents t'DP 11426 Section(c) Tabte(s) _ _ , Figure Change ( 1 Yes (X) No Titto of Changes. TIP Relief Valve ord Pressure Regulator Modification StaMARYt A contributing factor to tne reactor scram on April 10, 1990 was pneumatic flow passing through t he TIP Indexers which caused an unexpected rapid depressueltation of Primary Containment Pneumatic Supply (PCPS) Header. This depressuritation of the PCPb Header, which was isolated from its supply because of a RPS Bus f ailure, coveed the MS1Vs to isolate whien resulted in a reactor scram. It was determined that this rapid depressueltation of the PCPS header was a result of a low setpoint on 4 out of 6 outlet TIP indexor ret.'ef valves. The TIP indexers are purged with PCPS to prevent oxygen and moisture Lulldup in the TIP guide tubes. This EDP addresses espping of the indexer relief valves and reducing the setpoint on the TIP purge regulator to minimize the flow from the PCPS System. A catastrophic failure of the purge system was postulated . as part of the originot design basis for EDP 4940 and certain design provisions are included to prevent degradation of the PCPS by the fatture of the TIP purge piping. The
' TIP purge valve assemb4 which houses the pressure reguistor also includes a pressure retlef valve to protect the indexers from a pressure regulator ,
fatture.
4 .g .. n: s SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES bge 42
$AFETY EVALUATION SLAawtY Safety Evaluation No: 90 0053 UFSAR Revision No. 4 _
Reference Doculoontt ELp 10478 Section(s) Table (s) Figure Change [X) Yes [ ] No Title of Change: Deletion of Foodwater Controt System Reactor Water Level, Signet Discrepancy Annunciator Alarm SUhe WtY: This modifAcetion deleted a feedwater control system reactor water tevet signal, discrepancy annunciator alarm because, due - to its smalt deviation set point (+2") and-subsequent numerous falso annunciations. it has been determined that this alarm is not required. Thie alarm and its asecciated inputs have no safety function. However. the changes made by this modification do affact the UFSAR (Fig. 7.7 4) and operator response to postu'ated accident / transient scenarios. The accident / transient scenarios affacted by failure of this modification are loss of feedwater and f eedwater overf eed to the reactori both of which are already anal,yaed in the UFSAR. The operators will continue to have sufficient time to observe and correct reactor level- anomalies as 4 additional high/ tow le vel, alarms (*4"I exist. l l
*4
.s w' --
SAFETY. EVALUATIONS' ENQ1NEERING DESION PACKAGES Page 43 SAFETY EVALUATION SLBAMRY Safety Evaluation Not 90-0054 REV 1 UFSAI. 'ivision No. 4 3 Reference h nt EDP 11429 Section(s)' 6.2 / Table (s) j Figure Change IX) Yes ( ) No Title of Changes Modification of the High Point Vent on Division i RHR Return Piping StBamRY: This EDP modifies the'high point vent by removing one of the two vent ; solation valves, shortening the . remaining pipe spoots and modifying the sock-o.lst fi M et weld to reduce the susceptibility to vibration loading. The proposed tiesign meets the dual containment barrier design code requirement through the use of one isolation valve and a screwed on pipe cap. Tne 10 CFR 60 Generat Design Criteria 55 and 60 provisions for penetrations to primary containment are met through the use of a manual isolation val,ve and a screwed on pipe cap- on this LLRT vont connection. The isolation of reactor cool, ant pressure boundary design requirements are satisfied by the use of the RHR system. inboard containment isolation valve and the vent isolation valve. This modification has no effect on the venting operation. The alternating stress -levels will be -reduced -by a factor of approximately 3. .The-classification of the piping beyond the remaining vent isolation valve is i currently ANS, 831.1.1600p rating and win remain so in the revised design. An effect of removing this val,ve is a reduction in the adtinistrative controls required to. ensure vent closure following use since there is one less barrier between the' header piping and containment. The existing -administrative control.s on valve line up and cap verification assure leak tightness.
.. ....w..... ................................................................
t l r
SAFETY EVALUATIONS ENGINEERING DES 10N PACKAGES-Page 44 SAFEW EVALLMTION StaAMRY Safety Evaluation Not 90-0057 UFSAR Revisi m No. 4 Ref erence Docwet _EDP 10965 Soctlon(o)
- Tabts.! s )
Figure Change (X) Yes I 1 Mo Tit'Le of Changet RWCU Domineratirer Valves Control Air Source Change StaNARY: This modification changes : the source of cont'aot air for RWCU domineralizer valves 03300F150A/B. F152A/B, F153A/B. and F15AA/B f rom the Station Air System to the Instrument Air System. This change is being made because corrosion _ products caused by moisture from the Statien Air System have caused severat valve fattures. The Instrument Air System provides clean, dry air t o - t he valves.- This mooification removes the existing air fitters and moisture separators, caps the original statisi air supply line, and ties the valves into the instrument air supply. This modification does nct change the vatves' function or operation. It dose-not'-change _the instrument air system's function as the additional air usage of the above valves- is intermittent and sman relative to the overau instrument air supply and volume. Overau performance of the above valves is enhanced by the use of clean, dry control air.
SAFETY EVALUATIONS ENGINE.ERING DEblGN PACKAGES Pa.go 45 SAFE 7Y EVALUATION e m Y Safety Evaluation No: 90-006g UFSAR Revistaa No. 4 Reference Document: EDP 11322 Sectlon(e) Table (s) 6.2-2 Figure Change IX) Yes ( 1 No Title of change: Primary Coolant Boundary Isolation Valve. B21F434 Replacement s
'NY :
J This modification replaces the original primary coolant isolation valve. 021F434 with another design. The original valve required replacement due to unacceptable valve seet teskage. This modification does not change the design basis, function, or modes of operation of this valve. The new valve specifications meet or exceed the specifications of the old valve and Fermi 2 pressure / temperature specifications, Physical dif f erences between the old and new valves ! valve body, material, weight, and center of gravity) have been evaluated. Substitution of SA-182F 316L for SA-105 carbon eteet in the valve body meets or exceeds original valve specifications. The extra weight (4 lbs.) and slightly taller actuator (outward center of gravity shift) of the new valve have been subjected to a piping stress analysis. The evaluation concluded that this valve replacement has no substantial, impact on the pipe or hanger stresses.
SAFETY EVALUATIONS :j
.ENClhEERING DESION PACKAGES Page 46 SAFETY EVALUATION SLAAWtY Safety Evaluation Not 90 0060 UFSAR Revision No. 4 Reforence Document: EDP 110S4 Seetion(s)
Tabte(a) Figure Change (X) Yes ( ) No Title of Changet Revision of Sefety Parameter Disptey System Curves SLAMARY: The- saf ety evaluation was written to address changes to _the Saf ety Parameter Display System (SPDS) curves. The curves were revised . to be consistent with revisions -to the Emergency Operating Procedures (8,0Ps). The specific changes to the E0Ps which caused. the SPDS curve changes were a , result - of a design change to the Standby 1,1 quid Control System, minor changes to the CDP support document calculations and a chenge to the instrument which monitors reactor pressure vesvot temperature, ihe SPOS only monitors plant paaameters and provides information for emergency
. response functions. It provides no controt functions. Att inputs of class 1E parameters are accomplished via quattfled signal isolation which eliminates the possibility of SPDS causing a failure to safety related instrument toops.
-o-
.e-s9.
b
.. SAFETY EVALUATIONS -
ENGINEERINO DESIGN PACKAGES Page 47 z SAFETY EVALUATION SulemRY - Safety Evaluation No: 90-0062 UFSAR Revision No. 4 Reference DocumeEt: EDP 6332 REV A Section(o' Table (s) Figure Change IX) Yes ( ) No Title of Change: Supplemental High Capacity Condenser Waterbox Drain System Instau stion StaaMRY - This ' modification provides a 1000 gpm condenser waterbox drain system to supplement the original 200-gpm waterbox drain system. The additional draining
; capacity wi n significantly- reduce the time required to drain down the.
condenser waterboxes resulting in decreened plant down time. This system - has the same purpose and f unction as the original - 200 gpm drain
.down system and is not required f or p1, ant operation or safe shutdown of the plant. Because this- is--a maintenance system, it .is isolated from the Circulating Water System during plant operation and, when in use, is lined up to waterboxes isolated from the Circulating Water System. Failure of this system results in Circulating Water System teak rates that are wett below the circulating water piping joint fatture ' leak - rate assumption in the Turbine-Buttding flooding analysis.
4%.. , _ , , , , % + - . -- ____-_____m--___-.
5- . 2 - %-. 2 a a - .,%.. ..4 ...m. . . .+..##., - t ...mm_. ...m.ii_2_.- .. . . s. 3 is !
.s'
~~
- - n - l
' SAFETY EVALUATIONS-ENGINEERING DESIGN PACKAGES-iPage 48 ,
SAFETY EVALUATION SLAAMRY Safety Eval,uation No 90-0066 UF!MR Revision No. 4 Ref orence Docuinent EDP 11325- Section(e) Tabte(s) Figure Change (X1 Yes ( 1 No Title of Change Modification to Diesel, Fire Pump Fuet 011 Tank SLAAMRY: This saf ety ' evaluation was written to address the replacement of a tank drain , plug with a nipple, valve, and pipe cap assembly tocated at the bottom of the dieset' driven . fire pump fust ott tank. The purpose of this assembly is to
- provide a means - to -drain sediment and water of f ' the bottom of the tank without causing unnecessary spiu age of fust cit.
- The addition of the assembly win have no bearing 'on the ability of the fuet
-- olt tank to _ maintain the f uel, for operability of the diesel driven fire protection pump. No equipment matfunction la created by replacing the tank's !
currently instal.1,ed plug with the new assembly.
-- ~ - - e -
a.
.o SAFETY EVALUATIONS ENQ1NEERING DESIGN PACKAGES Page 49 SAFETY EVALUATION SO4MRY Safety Evaluation Hot 90 0000 UFSAR Revision No. 4 Reference Document: EDP 11481 Section(s)
Table (s) s Figure Change (X) Yes ( ) Ho Title of Change: Control Rod Drive Temperature Recorder Rept,acement SLA4MRY: This modification replaces the present control rod drive temperature recorder, temperature scanner, and temperature recorder banks with one multibank temperature recorder and two 120 point multiplexers. The original control rod drive temperature recording system was not reliable and was difficult to road. The system did not provide individuat temperature point bypass or alarm reset capability for icmperature points stabilized above the CRO high temperature alarm setpoint. As a result, the contret rod drive high temperature alarm setpoint had to be raised above the setpoint currently recommended by GE in order to avoid nuisance alarms. The new controt rod drive monitoring system consists of a Tracor Westronics DDR10/MB recorder and two W120 multiplexers. The replace.ent ef the contret rod drive temperature recording system is intended to improve the reliability of the system and make it easier to monitor controt rod drive temperatures. This replacement will also al' tow the control rod drive high temperature alarm setpoint to be decreased to the setpoint recommended by GE. This should increase control, rod drive s e al, life and decrease control rod drive maintenance because it provides Operations with an earlier warning of high control rod drive temperature. The recorder and multiplexer units are part of the Reactor Manual Control System and, as such, have no safety functions. The function of the recorder and multiplexer units remains the same as originally designed and does not affect the performance of any equipment or component important to safety.
SAFETY EVAL.UATIONS ENQ1NEERING DESION PACKAGES Page 50 SAFETY EVALUATION S M Y Safety Evaluatton Hot 90-0090 UFMR Reviston No. 4 Re f eronce Doctanent : EDP 11631 Section(o) 6.2 Table (s) Figure Change [ ] Yes (X1 No Title of Change LLRT Test Connection Modification for Steam to HPCI Turbine Outboard Containment Isolation Valve "NY: This modification inssetts a nipple and seat welded cap in the va\ve body of E4150F003 betwee the valve seating surfaces to provide a test connection for LLRT testing. This win aMow the bor.ne t cavity area to be pressurized resulting in a reverse flow LLRT teakage test. (See SE 90-0091 for evaluation of reverse flow LLRT testing of E4150F003.) The test connection does not crects an interf erence with the valve's moving parts or seating surfaces and failure of the test connection wou\d be bounded by the UFSAR sma n line break outside containment accident / transient analysis. i l 1
9
.t,;
. yf*- !
- . }
1
' SAFETY EVALUATIONS- ,
ENQ1NCERING DESIGN _ PACKAGES Page 61 SAFETY EVALUATION 1RANARY Safety Evaluation No 0101 REV 1 UFSAA Revision No.- 4 Ref eronce Document : . EDp 9845 Section(s) Table (e) Figure Change (X1 Yes [-1 No Title of Change: Instauntion_ of Additionat General Service -Water (GSW) Outlet Hender Pressure Pneumatic Controtter , StamARY: This modification instaus an additional, pneumatic controuer to eliminate the excessive 1. pressure - drop across the General Service Water header - upstream pressure . controt volve. The new controner maintains the pressure ~ at the outlet-of the upstream pressure controt valve at 50 psig by controuing the. downstream pressure control valve. This modification also-increases the header pressure setpoint from 140 psig to 150 psig to agree with the UFSAR. This modification .does not af f ect - the ability to saf ety shutdown the reactor because the 'OSW system does not provide cooling for saf ety rol.ated equipment ' during postulated accidents. 4
- c I
l L
p
.c l 'o SAFETY EVALUATIONS t
EN0!NEERING DESION PACKAGES Page 52 ,
- p SAFETY EVALUATION SLAAMRY Safety Evaluation Not _90-0112 UFSAR Revision No. 4 j Ref erence Document : EDP 11790 Section(s)
Tabte(s) i Figure Change [X) Yes . [ ] No ! Title of Changes- Miscettaneous Reactor Water Clean Up (RWCU) Pump "D" Support J Systems Modifications StammRY: The purpose of this modification is tot
- 1. Provide additional support to RWCU Pump "0" suction piping to reduce-
-mechanical loads and thermat impacts on the pump suction nerrte.
- 2. Remove-th<, suction strainer for RWCU Pump "B" to significantly reduce radiation fields in the room. The strainer was originally instatted
-for startup purposes and is no longer needed. The crud trap created by the empty strainer body makes it the highest radiation source in
'the pump room.
- 3. Provide-flex hoses f or the cooling water lines to _the oil cooter on the - RWCU Pump "B" skid to facilitate equipment fit-up and minimize maintenance activities and exposure in the pump room.
4 Change the OA tevet designations f or RWCU Pumps A & B f rom IM to HQ because the pumps are in the non-0, seismic II/I portion of the system- , and do not have a safety related or technical specification function, 1 The function and operation of RWCU Pump "B" is unchanged by this EDP. The failure of components modified by this EDP is bounded by the high energy line break analysis. A1.1 plant safety systems functions._-remain- unchanged .and l' reactor water chemistry parameters are unaffected by this change, l
. . ' 'l D'.
SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES -i Page 63: SAFETY EVALUATION SUhAMRY l F Safety Evaluation No: 90-0146 UFSAR Revision No. 4 Referonce Document: 0201 Soction(s) 9.2 Table (s) 6.2-1 Figure Change ( 1 Yes (X) No
= Title of Change ResiduaI Heat Removat Service Water (RHRSW) Design Flowrate Reduction Evaluation S M Yt'
' The ' purpose Lof - this safety evaluation . is to evaluate the of f acts of the reduction s' the RHRSW system design flowrate. - As a result wf-deviation event-report DER 90-0352, the RHRSW system - hydraulic analysis . DC-0201, was revised.
to include - the ' head toss ~ across - restriction orifices R0 - E11650001 and R0 E11550002-which had previously been omitted in the-calculation. The results of this ' analysis show that- the 9000 gpm design flowrate cannot be maintained af ter approximately - 137. hours..- _ Since the - containment anal,ysis. assumes a constant
- ' RHRSW fi,owrate of 9000 gpm during thsi fiest 30 days after the accident, the; reduced flowrate was evaluated for its effects on peak torus water temperature and ECCS, pump availabte NPSH. The results' indicate that the peak torus water temperature is not' increased and the ECCS available NpSH is not_ reduced.
I LAs a result of this evaluation, it can be concluded that the reduced f1,ow rate ~ o f . t he RHRSW system does' not decrease the heat transf er capcity of the RHR , _ 'systom_below'the amount of decay heat produced. 'There is no phy.ieel, change to L ; plant; equipment and RHRSW pump , reliability 'and' torus integrity -- romain unchanged. i l 1 t . . . . -- .
c: g , s. SAFETY EVALUATIONS ENQ1NEERING DESIGN PACKAGES
= Page 64 SAFETY EVALUATION StBAMRY Safety Evaluation No: 91-0014 UFSAR Revision No.' 4 Reference Document: EDP 3395 Sectionts)
Table (s) Figure Change (X) Yes ( l No e i litte of Change: Emergency Dieset Generator (EDO) Fuot 011 Day Tank Local Level Cage Replacement StaaMRY: This change replaces the EDO fuet cit day tank- locat- level, differential
. pressure gage on att four day tanks with a direct-reading pressure page.. The existing-tow side tubing between the day tank and the instrument-rack la-used in - t he new design. The original high side tubing f rom the day tank to the
- instrument rock is capped and abandoned in - place. The 5 valve manif old used with the original differential _ pressure gage has_ been. removed. The dif ferential pressure cett'was replaced because the high side reference tog would'not remain flued due to leakage across the instrument manifold bypass
,bl,ock valves and the cou diaphrams. This resulted in fatso tevet indication.
The new design 'etiminutes the need f or the manif old_ and reference tog by utilizing a: pressure gage which measures levet based on the head of the oil.
'This change does'not.affoct the operation or function of the EDn fuel oil day
~
tank. The pressure gage and tubing configuration maintains the same quality levet end seismic classification as the originst equipment.
.........'i..................................................................i..
l' SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES Page 65 IMFETY EVALUATION todWARY Safety Evaluation Not- 91-0020 _ UFSAR Revision No.- 4 Ref erence Document EDP 4300 Sectlon(s) , Table (s) Figure Change (X) Yes [ ] No-Title of Changet Addition of Pump Discharge Pressure Cages to UFSAR Figures StAAMRY: This safety evaluation was written to justify adding pressure pages and their associated isolation valves f or the sump pumps in reactor building equipment 'i and f toor drain _ sumps 01101-D065. D073 .D074, 0075 and 0076 to UFSAR figures 11.2-16, sheets "_ l ' and - 2.
'The pressure gages . were instaued as a part of EDP-4300. -This EDP also added mesh screens to each corner sump pump intake and mesh sce ; ening to - the underejde of the torus area sump D060 sump grating.
However, these changes do not affect any UFSAR figures. The addition of discharge pressure gages. enhances the operation and reliability of the floor drain system by identifying any reduction in pump performance prior.to comp 1,ete pump failure.- The piping, volves. : and supports are designed to . seist.ic . category-- II/I requirements and to the applicable ASME - and ANSI codes. The addition of sump pump intake screens prevents the' sump pumps from
-inges ting : debris which could damage the pump or - prevent normat check valve operation. The addition of screening to - the grating for sump 0065 prevents
' debris - f rom f aning into the sump. - The floor drain system is not required f or.
the saf e.~ shutdown of : the - plant . However, during a seismic event, it v!M remain intact,and not render any. safety related system inoperabl.a. i l
(- .'. 4= ri r SAFETY EVALUATIONS ENQ1NEERING DESIGN PACKAGES Page 66 , SAFETY EVALUATION InasMRY Safety Evaluation No 91-0023 UFSAR Revision No. 4 1 Reference Document: EDP 10126 Section(s) Table (s) Figure ( W, (X) Yes ( ) No . Title of Change Addition of a Third Decenting Pump to UFSAR Figure 10.4-6 l i StaaMRY: This safety evaluation evaluates a revision to UFSAR figure 10.4-5 to include the instauntion of Circulating Water Reservoir Decanting Pump 83. The pump (10000 gpm 9 16ft. head) and the motor (460 v. , 76hp. 8 800 rpm.) have been - instaned per EDP-10125. The pump power ' supply w1M be' instaued under a different EDP at a future date. The addition : of an extra decent pump will anow the circulating water system inventory to . remain balanced and anow system chemistry to remain within acceptable limits with one decent pump out of service. Operation of three decanting pumps .witt attow for optimal reduction in circulating : water pond temperature during the eummer by increasing the amount of water that can be recirculated between the pond and Lake Erie. The-decent pumps are not required to perform any safety related function. The mechanical insta uation of the pump meets-the requirements ~of ANSI D31.1. With the operation of att three pumps, total decent flow will not exceed the NPDES i- permit of 45,000,000 gattons per day. EPA and Michigan Department of Naturat !. Resources (DNR) criteria for total residual chlorine and radioactivity are met l or unaffected by the instauntion of a third decent pump.
- END OF EDP SECTION l
't *1 e
FERMI 2 SAFETY EVALUATION
SUMMARY
REPORT 1990 POTENTIAL DESIGN CHANGES 4
3.E
.;D i
SAFETY EVALUATIONS POTENTIAL DESIGN CHANGES Page 1 SAFETY EVALUATION SLAAMRY . I Safety Evaluation No: 88-0046 REV 1 UFSAR Revision No. 4 Reference Document: POC 8121 Section(s) 6.2 .i Tabte(s) :
.l l
Figure Change ( 1 Yes (X1 No l l i Title of Change Main Steam Isotation Valve Leakage Control System Setpoint -- Changes SLAAMRY: Certain nominet trip setpoints (NTSps) in the main steem isolation valve leakage- control system; (MSIV/LCS) require changing in- order to obtain consiatency with the General, Etectric eetpoint methodology. Thia methodologry provides the basis for setpoint~ sel,ection involving instrumentation having
-Technical Specification survetM ance requirements. The new setpoints are more conservative than the previous values. 1 The ability of the MSIV/LCS to perform its function is enhanced by those changes since the margin f rom setpoint to analytical,1,imit le increased. The setpoint changes can'not contribute to an accident. Design calculations show the chosen setpoints provide adequate protection against. exceeding anatytical limits and against spurious trips. No equipment was changed. Interaction of equipreent functions. as controued by these setpoints, is basically unchanged. 5 I
l l 1 l
t* SAFETY EVALUATIONS POTENTIAL DESIGN CHANGES Page 2 SAFETY EVALUATION SLBAAARY Safety Evaluation No 88-0119 UFSAR Revision No. 4 Refarence Document: POC 9090 Section(s) Tablets) Figure Change (X) Yes ( ) No Title of Change North and South Reactor Feed Pump Drip Drain Isolation Valves Removat and Piping Reroute StASAARY: This modification removed the north and south reactor feed pump drip isolation valves and rerouted tha south reactor feed pump drip drain piping. The isolation valves were removed because the possibility exists that if the valves were closed, the drip cavity would fill with seat water. It is then possible for the water to enter the bearing oil area and contaminate the tubricating cit. The associated drip drain piping of the South reactor food pump was rerouted from an equipment drain to the floor drain system. This was done because the equipment drain system is not equipped to handle potentially oily water, The floor drain system utilizes an cit / water separator before the water is processed in the radwaste system. The design conditions for the delp drain piping were revised from 950 psig 9 430 degrees F to 14.7 psia 0 212 degrees F since the piping and the pump bearing hub are open to the atmosphere. This modification insures the reliable operation of the reactor feed pumps by decreasing the probability of water contaminating the tubricating oil. Rerouting of the drip drain piping decreases the potential of oil entering the equipment drain system, thus improving the operation of the radweste system. The subject piping and valves are not covered by the technical specifications. Therefore, there is no impact on the technical, specifications.
. . . . -- . . ~. . .- . - .-.
..g
_ (4 - i' SAFETY EVALUATIONS POTENTIAL DESIGN CHANGES Page 3 SAFETY EVALUATION SLANARY i Safety Evaluation No 68-0139 UFSAR Revision No. 4 Reference Document POC C127 Section(s)
~
Tabte(s) Figure Change IX1 Yes [ ] No Title of Change: Instauntion of Pneumatic Bias Relay StnNARYs A pneumatic bias relay was instaued in the control toop for reheater seat tank (RST) . tevet _ to f eedwater heater #6 North. The RST flows .to both heater N6 North and #6 - South, but there is only a single level controner. Due to inadvertent process / hardware - dif f erences, the flows are not equal. The bias relay enabi,es the two paths to be balanced. Falture of the blas relay. in _ any mode , . i s bounded by " loss of feedwater heating," in the UFSAR. -If the rol,ayJ f ailure stops the RST flow to feedwater heater N6 North, extraction steam from the HP, turbine win increase. Finat feedwater temperature will remain essentially the.same. On the other hand, if the relay f atture causes the control valve- to' futty open, most of the RST will preferentiauy flow to feedwater . heater .NG North, extraction steam wiu decline; final feedwater temperature wi n remain essentla n y the same. t w ,,e -
. .~, . . _ . - . - _ - . . - - .. - .~. - _ - . . ,
a,; SAFETY EVALUATIONS POTENTIAL DEB 1GN CHANGES Page 4 ! SAFETY EVALLMTION StaamRY~
' t Safety Evaluation Not 88-0219 UFSAR Revision No. 4 Ref erence Doctament POC 9698 Section(s) 8.3 Tabte(s)
Figure Change (X) Yes [ .1 No Title of Change: 260/130V Battery Charger Undervoltage Alarm Setpoint EAANARY . The' tow voltage alarm for the division I and !! essentist switchgear systems (ESS) and balance-of-platit (BOP) battery chargers was increased from 124.5 + 0.5v to 128.5 + 0.5v. The change made was only to increase the low vol,tage setpoint so that the atarm signet witt initiate ut higher battery terminal voltage and detection of any division-.-. ! and II ESS and BOP battery charger malfunctions wil,1, be much sooner. The new - voltage alarm setpoint of ' 128.5v is consistent with the
= minimum - acceptable voltage ~ of 2.13v per cett as specified in - the Technicat
; specifications.
e W i n
-c-- -- -
. ._m._ _ _ . _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _
r: - .
*e SAFETY EVALUATIONS POTENTIAL DESIGN CHANGES.
.pago 5
- SAFETY EVALUATION StaAAARY -
Safety Evaluation No 88-0239 ~ UFSAR Revision No. 4 Reference Document POC 8823 REV A Section(s).. 9.A.6 Tabte(s) Figure Change I 1 Yes (X) No Title of Changes Fire P o p Betteries Specific Oravity Change SGAAARY: This modification changed - the dieset-driven fire pump starting -24v batteries specific gravity requirement from 1.200 to 1.235 and the overatt battery bank voltage from 24v to 26.2v. Since tho ' dieset-driven fire pump - has been placed in . service, the battery system has failed, in severat instances, to start the pump. The root cause, as determined in deviatinn event report DER 87-0330, was that the specific gravity value of'1.200, as specified in Technicet Specification 4.7.7.1.3 was too low to. ensure that the battery would'be sufficiently charged to consistently start
'the dieset-driven fire pump. The increase in specific. gravity requirement from 1.200 to.1.235 provides sufficient charge for the 24 volt battery to start the dieset-driven-fire pump. The increase in minimum specific gravity requiroment,.
from 1.200 to 1 234 inc rea n t.,s the- acidity of'.the. electrotyte and hence increases ' the battery minimum state of charge. .As a result of this, the dieset-driven fire pump has started after cranking for oni,y-two seconds.
SAFETY EVALUATIONS
-POTENTIAL. DESIGN CHANGES-Page 64 SAFETY EVALUATION-StAARRY
- Safety Evaluation Not 88-0242 RCV 1 UFSAR Revision No. '4 Ref erence Doctament : PDC 9923 Section(s)._ _ ,
Table (s) 7.3 F1gure Change ! 1 Yes- tX) No Title of.Changet.- Core Spray -Sparger 01fforenttal Pressure Switch- Trip Setpoint and Range Change i SuhAERY: --
. The: core spray sparger dif f orential, pressure switch was recalibrated and the :
.- nominal'._ trip.setpointc(NTSP) was changed from 0.5 paid-to 0.2 paid increasing.
!To maintain- conservatism. .ithe core . spray sparger dif f orential. - pressure -' switch L E21NOO4A/B range was changed f rom ~5,' +5 psid to .-7. +2 paid.
Tho' nominal trip setpoint (NTSP) change is in the conservative. direction. ,The , associated- dif f erential pressure _ switches provide an alarm function'only, and - have no ' auto-interlocks - with any . other . syntems. Also, this measurement is capabte of a break detection f unction at al.1 power and flow > tevels. _-: Based on
- operating. data taken at_above 80% reector power and 90% core flow, the initial
- condition f or :a postulated ~ sparger break is -3.3 psid (normal plant operating- .
;- condition). Based on the >CE ' design Especification, .the change in -the'
-'dif f orential . pressure . of +4 - paid - is 1 expected. when a : break . in the sparger 1
occurs. Using this data and allowing for-expected instrument drif ts design -
- calcul,ation - (DC)- 4584 has been revised. 'The results of the - DC 4584 revision
-indicates the need to lower the setpoint to make it more conservative.
y M-?
'.: I
.- J, .
- i SAFETY EVALUAYIONS .;
POTENTIAL DESIGN CHANCES ' Pege 7 - SAFETY EVALUATION SLASMRY
.j '
Safety Evaluation No: 89-0016' UFSAR Revision No. 4 Ref erence Doctament:-- POC 9826 Section(s) i Table (s) ; i Figure Change [X) Yes I 1 No i Title of Change: Romete Shutdown Pane 1 Human Enginsering Def1cienc1es Resolution i SLAAMRY: .. l. The Detalted Controt Room Design Review (DCRDR). Team was established to fulfi n the requirements of-NUREG's 0700 and 0737 Supplement 1 to identify and provide ; suggested improvements to the control room and remote shutdown panet that would strengthen the - man-machine ' interf ace. As a resul,t of this review, human engineering deficiencies --(HED) were identified at the remote shutdown panet. This minor modification was issued to resolve the HEDs. The modifications to the. remote shutd;wn penet are to enhance the 1,ogibility of componenta on the pane 1 and provide consiatency in design with the main control, room , penets , i The modifications' are . designed to help the operator to- identif y equipment- controts more readily.,and provide the . operator ~ with accurate equipment = status (i.e., addition of "PSIG" label,s to RPV and drywou pressure - cl
- indica t ors ) . .These enhancements will reduce the probmoltity of operator errors dueto- inaccurate interpretation of- displayed plant data or incorrect l identification of plant control,s which could adversely affect plant safety, i
i 3
.. . g-SAFETY EVALUATIONS 1
POTENTIAL DESIGN CHANGES Page 8 SAFETY EVALUATION WaemRY - Safety Evaluation No 89-0016 UFSAR Revision No.- 4 Referance Document: POC 8475 Section(s) Table (s)- Figure Change (X) Yes [ ] No Title of Change: Feedwater Flow and Recirculation Motor Generator- Set Controller Labeling Changes SLEAMRY:- Two- new labels were instaued on the f eedwater _ f tow controtters. Six tabets with revised engravings- were instaued for the feedwater controu ers- and indicators on the associated control room. panels. Two new labels were instatted on the recirculation motor generator (MG) set controtters. The new labels for the f eedwater controuers use more consistent nomenclature than the previous labels. The new tabels for the recirc MG controMers anow tactito feedback distinguishing the feedwater flow and recirc MG set controuers. The new s ele are designed to enhance the operators ability to correctly identify these controllers and- indicators on the- control board. The installation of these labels has no ef f ect on the manual or. automatic startup and operation of any plant systems. No new equipment is added and no setpoints are changed. END OF PDC SECTION-
FERMI 2 SAFETY EVALUATION
SUMMARY
REPORT 1990 PROCEDURES, TESTS, AND EXPERIMENTS 1
4 SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS Page 1 SAFETY EVALUATION DG4MRY Safety Evaluation No: 09-0122 REV 1 UFSAR Revision No. 4 Ref erence Document : DER 88-1801 Section(s) 11.43 12.2 Table (s) FiDure Change ( 1 Yes (X1 No Title of Change: C1,arification of Turbine Building Dampers Functional Description Suh4RRY: The wording in the referenced UFSAR Sections states that some Turbine Building dampers are used to automatically isolate the Turbine Du11 ding from the environment. De-energitation of the supply / exhaust fans will result in closure of backdraft dampers which removes the exfiltration mechanism for the Turbine Building atmosphere. This type of isolation is sufficient for the Turbine Building Ventilation System because this system is required to function under normat operating conditions only and is not spa-ifically designed to operate after a design basis accident. Therefore, the existing UFSAR wording pertaining to automatic isolation of some Turbine Duilding dampers can be removed. The backdraft dampers provide the same func, tion as the Turbine Duilding dampers which are designed to automatically isolate and are closed by fan trips. However, as stated above, the isolation function served by these backdraft dampers is not required during accident conditions. l \ I e
/ ' l' SAFETY EVALUATIONS PROCEDURES. TESTS, AND EXPERIMENTS Page 2 MrETY EVAltlATION "NY Safety Evaluation No 89-0162 UFMR Revision No. 4 Reference Document: DER 88-0241 Section(s) 5.2 Table (s) Figure Change ( 1 Yes (X) No Title of Changes Mechanical Stress Improvement Process (MSIP) Implementation e m RY: This safety evaluation justifies the use of MSIP to complete the mitigation actions for prevention or minimization of intergranular stress corrosion cracking (IGSCC) in t he susceptible etainless stent reactor coolant pressure boundary piping and reactor pressure vesset nozzles. MSIP improves the residual stress distribution in a weldment . MSIP is perf ormed by installing clamp rings to the outer surfsce of the pipe adjacent to so weld. The clamp rings are brought together by stud tensioners or box pre.sres. The clamp rings are tightened to slightly contract the pipe watt to a eintrotted, predefined condition. After comoval of the clamp rings, a locat, r,ormanent reduction in the outside pipe diameter results in the desired rircumferential residual stresses. During the first ref ueling outage. MSIP was perf ormed on twenty-seven susceptible welds. The implementation of MSIP does not change the design, function, or operation of t he affected systems. The application of MSIP has no effect upon the operational parameters of the components such as temperature, pressure, materlat selection, or design attowable stresses or loads. EPRI NDE Center tests have confirmed the analytical calculations which predicted that the process is incapable of propogating existing cracks and, analyticany, the process has been shown to be incapable of initiating new cracks. Implementation of MSIP decreases the potential for cracking in the piping being treated. The implementation of MSIP does not affect any other safety related system or component which is required to have an active function. l. _ ~ ._ _ __ _ _ _ _ _ _ _ _ _ - - - - - _ _ _ - - _ - - _ - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' " - -
' ',1 3 SAFETY EVALUAT1DNS PROCEDURES, TESTS, AND EXPERIMENTS-Page-3
, SAFETY EVALUATION SLAAMRY Safety Evaluation No 89-0191 UFSAR Revision No. 4 Reference Document: DER 84-2039 SectionOt) A.8.12 Tabte(s) i Figure Change (X) Yes I 1 No Title of Change: Celticality Monitora f or Storage of New - Fuet in the Fuel
]'
-Poot 1RAAMRY:
Since. the fuel poot is being used to store new fuet, detectors D21-N116 rnd . D21-N117 have been designated as the criticality. monitors foe the- f uel pool, storage area.- T us. safety analysis. assesses the compliance of these two-detectors with Reg. Guide 8.12.- j Reg. Guide -8.12 .provides guidance on criticatity eterm systems that are acceptabl.a . to the NRC staff. It states that the system is acceptable if it mee t s ' the guidance of ANSI standard ANSI /ANS-8.3-1986 and that each area is covered by two-detectors (as required by.10CFR70.24(a)(1)). The detectors meet the requirements of ANSI /ANS-8.3-1986 in tnat, for the range of detector al, arm setpoints aM owed by.the technical specifications, the maximum ' distance On the ref ueling floor to - the ~ detectors- is less than the maximum aMowed by the standard. The redundancy requirements of 10CTR70.24(a)(1) are met in that there are- two = detectors used as criticality monitors. This monitor 13g system is - designed .so that it alarms for any system matfunction or = toss of power. i The system is periodiceMy tested to ensure its opernoitity. 4' The two detectors used for spent fuel pool criticality monitoring are identical in make and model to the sing 1,e detector used for criticality monitoring in the new fuet vault. The reliability of this system witt he higher because two monitors are used.- _ .. ___.___________________________a__ _ -
- . - . ~ . ~ . _ .
. .. _ . . - - ~ - - - _ _ - - . . _ . - - . . - ~ . . . . _
a.
- J SArtTY EVALUATIONS PROCEDURE 8, TESTS, AND EXPER1W.NTS l Page 4 1 ,
! SAFETY EVALUATION Sl4AMMY 1 Safety Evaluation Nel 90-0036_ UFSAR Revision No. N/A Ref erence Doctament NPP 23.420 _, Section(s) Table (s) Figure Change ! 1 Yes (X) No Title of Changel *RHR Complex Heating and Vent 1\ation* Procedure Change StatMRY: This procedere was revised to inctvde guidance on the requirements necessary to maintain the ED0s end associated equipment operabte during Residual Heat Removat (MHR) Complex damper maintenance. Working on the RHR Complex dampers in accordance with the guidelines af tne subject procedure wilt not prevent the HVAC . sys'.em from maintaining room temperatures whithin their design basis requirements. Beismic coquiremente - wiu be met and not impact the failure probability of the dampers. Any damper fatture would be bounded by the analysis in the UFSAR which indicates that if temperatures should rise / fan to alarm tevote, operators would take necessary actions to reatore temperatures to , acceptable levote. i h 1 F 1
1 S SAFETY [ VALUATION $ PROCEDURES, TESTS, AND EXPERIMENTS Page 5 SAFETY EVALLMT!uN SMY Safety Evaluation No: 90-0038 UFSAM Revision No. 4 Re f erence Doctament : N/A Section(s) 12.31 13.1 Tabtr(a) _ Floure Change IX) Yes ( ) No Title of Change Consolidation of the Radiation Protection Manager posi; ion SMY The current responsibititles and f unctions of the Radiation Protection Manag*r (RPM) position have been added to the Superintendent . Radiation Protection (RP) position and - the- RPM position- is to be - eliminated.- Existing responsibilities and f unc* fint, as welt es the current reporting tevet of the Superintendent . RP. position have been retained. Required training and . qualifications for the RPM and Superintendent RP have been combined such that
. the more restrictive requirements f or both pcsitions are retained.
- Contotidation of the RN position within _ tho eAioting Superintendent - RP position is intended to accomplish the following:
- 1. Improve reporting by the RPM funntion to the plant manages *nt which in consistent with NRC Regulatory Guides 8,8 and 8.10, 2, Strengthen the accountability of the RPM f unction f or the priorities, resources, and results of the Radiation Protection Organismilon which is consistent with NRC Regulatory Guides 8.8 and 8.10,
- 3. Simplify the organization and thereby provide for more efficient utitiration of resources.
c .
,. .c , , ., ..-.. . , , - . - . . . . ,e. ..-,--,,..,,.....w....-,. . , _ , , , . . _ . _ _ _ . - . - , - , . , - . . . - . - . . _ . , . ,
SAFLTY [ VALUATIONS PRDC(DURES, TLSTS, AND EXPERIMENTS Page 6 CAFETY EVAlt1ATION Sl&3ARY Saf ety Evntuotism Not 90+0043 REV 1 UFLAR Revision No. 4 Reference Doctament: UFSAR Sectitm(s) 7.6 Tab \e(s) 6.2-2: 6.2-13 5.2-1b 7.3-93,7.6-2 7.6-9 l'1gure Change tX1 Yes ( ) No Title of Change: Correction of Miscotlaneous UFSAR Discrepancies SLMMRY: The purpose of this safety evaluation is to eva\vate the correction of miscel\aneous discrepancies in the UFEAR discovered during the Tectmical Specification Improvement Program audit of technical specifications 3/4.2,2. 3/4.3.2, 3/4.6.3. and 3/4.7.4 and during the preparation of DC 4993. The discrepancies includedt
- editorist corrections
- addition and detetton of votves
- correction of valve isolation logic relationships
- correction of vatve, drawing, and reference identification numbers
- correction of clarification notes No physical changes to the plant were made by these revisions. The postulated accidents and the accident analysis described in UFSAR chapters 6 and 15 are not affected by these changes.
1
9 s bML1Y CVALUATIONS PROCEDURES, TESTS, AND [ APf RIMENTS Vsge 1 SAFETY LVALL1ATION DLAARRY Sefety Evaluatton Not D0-0047 UFSAR Hovision No. 4 Heferonce Doctanent: DLR 90-0107 Section(e)
~
Tabte(e) Figure Chance IX) Yes t 1 No Title of Chan,e Technicet Specification Fire Door Documentation and Evaluation SLAW AY: This safety evaluation identifies doore 13-6 and 114 0 , in . their present locations, se the fire rated en embly protecting the opening between the Tagging Center and the Control Room and the opening botween the Computer Room and the Turbine Building, respectively, Door R4-0 le a class A fire door. Door T3-6 is considered as capable of meeting the acceptance criteria of UL 100 for the required 1 1/2 hour rating and has been granted a deviation for use as a fire barrier for the Control Room, As auch, both doors meet the rated boundary require.nents of the UFRAR Fire Hazards Analysis. ( 1 l
*, 0
- SAFETY CVALVATIONS PROCEDURES, TE$18, AND EXPERIMENTS P6pe 8 l
SAFETY EVALUATION tRaamRY Safety Evaluation No 90-0068 UFSAR Revision No. 4 ! Ref erence Doctanent: DER 89-0736 Section(s) _9A.6 Tobie(s) l Figure Change ( ) Yes (X1 No 1
- Title of Change - Attornate and Equivalent Means of Supervising Fire Door and Caseous Bour.dary Door Position
)
1RasmRY: l This saf ety evaluation revises the UFBAR description of fire door supervisory i alarms to ' reco0 nite methods of monitorinD fire door positions equivatont .to - those presented in the UF&AR and Appendix A to BTP ABB 9.6.1, dated August 23, 1976. Appendix A of. BTP ASB 9.6.1, dated 8/23/76, states that fire doors be 'normatty closed and tocked or alarmed with alarm and annunciation in the contret room." The present NRC guideline, CMEB BTP 9.b-1, rev. 2, dated 9/81, includes ! guidelines which verify fire door position by checking the position of each tocked closed door every seven-days and checking.the position of each unlocked door every twenty-four hours. This guideline also attows the monitoring of. fire
~
door position f rom a ' " constantly at tended" tocation. The fire doors at Fermi
.are monitored in accordance with the above guidelines,'which etto were included -
in Technicat Specification 3/4.7.8. Appendix = A' of L BTP ASB 9.6.1 dated 8/23/76 does not address gaseous boundary doors. However, CMEB BTP 0.5-1, rev. 2, doos state that gaseous boundary doors be self closing and electricatty supervised. The Fire Protection Program .at Formi' does not commit to the requirements of CMEB BTP 9.6-1 and it does not-
-include procedures for monitoring gaseous boundary door position. Att gaseous boundary doors--at Fermi are normatty closed and are provided with some method of. monitoring door position. These methods include unlocked doors alarmed in--
the control room, locked doors starmed with sect.rity, weekly checks of locked closed doors, and deity checke of untocked doors.
, , . . . . , , . , , - . - , - - - -- , . , . - , . . - . . . - . - , , . . . ;-. _ .. ,, . . _.-.--.,___.a.--- -.
.~. _ _- _ _ . _ . = _ . _ - - _ _ - . _ . _ _ . - _ - _ _ _ _ _ _ _ - ~ . . . _ .. _ .__ ._
1 e, I4 SAFETY EVALUATIONS PROCEDURES, TE$TS, AND EXPIRIMENTS Page 9 Saf ety Evaluation 14o. 90-0058 (continued): The above methods provide e reasonable tevet of assurance that fire doors and gaseous boundary doors witt remain in the proper position and be discovered promptly if out of pJsition. The f unction and re\ lability of the fire doors remains the same as bef ore and the changes to the UFSAR reflect the present p\ ant configuration. 0
+, , - . - - - - - - - - - -
. . , . , , . - , , , - - , , -n.,_----.a, ,- . , n - -, --
e SAFETY EVALUATIONS PROCEDURES. TESTS, AND EXPERIMENTS Page 10 FAFETY EVALUA110N SLleMRY Safety Evaluation No 90-0064 UFSAM Revision No. N/A Referonce Document: NpP 23.108 Sectionta) Table (e) Figure Change ( 3 Yes (X) No Titte of Change Plant Operation with Foodwater Heater Extraction Steam Isolated St.W@Y: This evaluation was performed to change procedure NPP 23.108 to attow continued operation with extraction steam isolated to feedwater heater (s) provided feedwater temperature remains above 370 F. This safety evaluation has evaluated the consequences of operation with an extraction steam line isolated to ensure that the applicable UFSAR analyses are still bounding. The loss of f eedwater heating associated with a sing \o beater is an analyred event. This analysis ensures that the worst single falture limit. n 100 F reduction in feedwater temperature. is not exceeded. The isolation of extraction steam will not reduce the reheater bypass flow nor increase feedwater temperature change beyond the 100 F used in accident analysis and thus, can 'ot cause a pressurization event of greater magnitude than the previously .alyzed events. Since the proposed change has no effect on the pressurization and toss of feedwater heating ownts, the existing UFSAR analysis is still bounding. The operating MCPR limits as specified in section 3.2.3 of the Technical Specifications are therefore atitt bounding and hence the margin of saf ety as specified in bas 49 of Technical Specifications is not
, reduced.
- . - . - - - - . - .. - ~ .. ~ ~ , - - . - - - ~ - .-- - - - ~ . . - -
- e. .
SAFETY EVALUATIONS-PROCEDURE 9, TESTS, AND EXPERIMENTS Page 11- ] SAFLTY EVALUATION managny Sefaty Evaluetion Not 90-0072 UFSAR hevie1on No. 4 Reference Documents D8TF N21-019 I Section(s) 5.63 7.63 7.7 l Tabte(s) ) l l Figure Change i 1 Yes IX] No l
~r
= Title of Change ' Feedwater System UFSAR Editorial and Minor Corrections 1- !Kate%RY:
This safety evaluation justifies editorist and minor corrections to the subject sections of the UFSAR that describe the feedwater system. The discrepancies were discovered during the. Design _ Basis Task Force review of the system. The changes concern Foodwater piping site description FW controt system function section typographical errors FW controt system power source discrepancies RPV~ water level, signal processing FW flow signal processing , Reactor water tevet trip of the reactor feedwater pumps Turbine driven FW pumps controt speed timiters Nuclear boiler system setpoints text reference There are no changes to - the design, f unction, or operation of the feedwater system. These ' changes correct the description of . the feedwater system to reflect the actual plant configuration and, as such, do not have any impact on plant safety.
.......... ................s..... .............................................
I n , , -m- w + , - o , . , , v n ,-_.,e----, , -- -,-..,v,- - - - - .n- ----~n.-- - -
~~-n-- r- - - - - + ~c' - ~ ~ ~ ' ~ ' ~ - ~ ~ ~ - = = ~ ~ ~
i . l l' CAFETY EVALVATIONS PROCEDURES, TESTS, AHO EXPERIMENTS Page 12 SAFETY EVALUATION SMY Safety Evaluation No: 90-0076 UFSAR Revision No. 4 _ Reieronce Document: Seetion(e) 8.3 _ Table (s) _ Figure Change ! ] Yes (X) No Titt,e of Change Reactor Protection System M3 Set Power Source Correction in the UFSAR StH MRY: The purpose of this safety evaluation is to evaluate a revision to the UFSAR that property identifies the normat and alternate power supp\les of the Reactor Protection System (RPS) MQ sets. UFSAR section 8.3.1.1.10 origina1\y stated that the 480 VAC switchgear provides a direct power source to the M3 sets. LCR 90-104-UFS corrects UFSAR section 8.3.1.1.10 to indicate that the direct primary power sources are MCC 720-40 and MCC 72E-60 and the second6ey power sources are hCC 72C-2D and MCC 72F-4B f or M3 sets A and D, respective \y. This revision wiu put UFSAR section 8.3.1.1.10 in conformance with the at-busit condition of the plant and with UFSAR figures 7.2-2, sheet I and 8.3.-6. This revision does not a\ter the functional or operational characteristics of the 400 VAC distribution system or the RPS and no new componente have been added to either the RPS or the 400 VAC distribution system.
0 4 SAFET) EVALUATIONS PROCEDURES TTSTS AND EXPERIMENTS Page 13 l SAFETY EVALUATION StamARY Safety Evaluation Not 90-0076 UFSAR Revision No. N/A Refarence Document: NPP 46.137.001 Section(s} Tabte(s) Figure Change I 1 Yes IX) No Title of Changet Respan Transmitter for Fuel Poot Levet stamWtY: This safety evaluation addresses a procedure change to respan/zero shift a fust poet levet transmitter. .This altows monitoring of.fust pool water levet.during refueting. The procedure modifies . the air pressure to the reference tog of a tevet transmitter to cause a range shif t of the instrument so that the fust pool levet up to the top of the skimmers can be monitored. The new pressurization value does not pose any t hroc.t to piping integrity because the required change ranga (meximum 29.9 peig) in significantly less than the design piping pressure of 1276 psig. Removat or failure of the floodup level instrument during refueling will not affect refusting interlocks or cause 's loss of required indication that could not be obtained by visual means. The floodup indicator instrument range will, only be af f ected during reactor pressure vesset flood up conditions, thereforo no change will occur to accident or normat shutdown monitoring functions. 9-
. y ,e-- , - . e -m - , - ..--.,y ,-- . , r-g - . + - r, ~- r- c y . %, - - . - w ..-, ,-wc-.
SAFt:TY EVALUATIONS PROCEDURES. TESTS, AND EXPERlhAENTS Page 14 SAFETY EVAlliATION
- N Y Safety Evatuation Not 90-0077 UFFAR Revision No. N/A Reference Document: HPP 46.137.002 Section(s)
Table (s) Figure Change ( ) Yes (X) No Title of Change: Conversion of Core Plate D/P Instrument to Ref usting Lovet Instrument SLM ERY: This saf ety evaluation addresses a change to procedure NPP-46.137.002 which will modify the function of the core plate D/P instrument so that it can serve as a reactor water levet instrument during refueling conditions when it to not required f or core plate D/P indication. This will provide a redundant levet indication with the floodup levet instrumentation and will attow removat of the floudup instrument for maintenance without a total lose of indication. The core plate D/P monitoring f unction does not provide any storm or trip functions and is not listed as safety-related power generation display instrumentation. The function of this instrument is to provide indication of the hydraulic perf ormance of the core internal components. During ref usting conditions, this dif f erentist pressure is below indicating range since core plate flow is extremely low. Since no safetv function is served and the indication is below scale, there will be no imptet by removat of this function. Based on the above, no safety significance wilt be associated with conversion of this instrument during refueling conditions.
SAFETY EVALUAT!DNS PROCEDURES, TESTS, AND EXPERIMENTS Page 16 SAFETY EVALUATION MA&RRY Safety Evaluation No 90-0085 UFSAR Revision No. 4 Ref erence Docusient : UFMR Ch. 13 & 17 Section(s) 13.11 17.2 Tebte(s) Figure Change tX) Yes ( ) No Title of Changet Nuclear Quality Assurance, Plant Safety. Nuclear Security. Nuclear Services, and Radictogical Assessment Organizational Changes SUWKRY: This safety evaluation ovatustos the fottowing orges.irat ional changes to be incorporated into the UFSAR:
- 1. Nucienn Quality Assurance reports to the Director Assurance.
Nuclear Quality
- 2. The Director - Nuclear Quality Assurance, Director - Plant safety.
Director . Nuclear Services, Director - Nuclear Security, and the Sonice Engineer . Radiation Assessment report to the General Director
- Nuclear Assurance.
3. Thw Gensett Director - Nuclear Assurance reports to the Senior Vice l President - Nuclear Generation. 4 Nucione Security reports to the General Director - Nuck.e Assurance. 6. Nuclear Services reports to the Generat Director . Nuclear Assurance. Despite the change in the reporting relationship described in the UFSAR. the QA program is in conformance with Reg. Guide 1.33 (Rev 2). "QA Program Requirements (Operations)" and ANSI N18.7 - 1976/ANS 3-2 as committed to in the UFSAR Appendix A and 10 CFR Appendix B. Technical Specification 6.2.1.C. It is also in conformance with
., t-
~. 1 SAFETY EVA60AT10NS PROCEDURES TESTS, AND EXPERIMENTS Page 16 a d SAFETY EVALUATION stanMRY P Safety Evaluation Not _90-0093 UFSAR Revision No. 4 t Reference Documents DER 83 0060 Section(s) 7.6-Tabte(s) Figure Change ' L 3 Yes (X) No
. Title of Changet UFSAR Section 7.0.1.4.2.6 Rewording 9tteMMY:
-Section 7.6.1.4.2.6 of the.UFLAR describes the Engineered Safety Features (ESF) status penet. This section was revised to delete any references to ' Technical Spoolfication Violations". This makes Section 7.6.1.4.2.6 agree with 5
Regulatory Guide 1,47 comptierce criterie. The ESF ' status penet is a humar, frictors aid to assist the operator (beyond administrative measures) if ' both traint of diverse, redundant ESF systems are rendered inoperable. This panet has no contret . functions and its indicating circuits are not linked electricatty or mechanically to any plant systems. Transient and accident analyses do not take credit for the ESF status panet. f l
.-,_..--..,m, .m..,- .,,m_ . _ , , , , , . . . . . , , ,m, ...,,...~.v..,~m, ...,,,m._ . . . , , . , . . , , . . , . . . . - . - . . . . . , . . - , . . - ~ . . , , ,
SAFETY CVALUAT!DNS PROCEDURES. TESTS, AND [XPERIMENTS Page 17 SAFETY EVALLATION ESAMRY Safety Evaluation Not 90-0110 UFSAR Revision No. 4 Reference Documentt DER 90-049829 Section(s) 3.33 9.13 12.1f 815.7 Table (s) i Figure Change ( 1 Yes IX) No Title of Changet WisceManeous UFSAR Changes to Correct Discrepancies identified in DER 90-0429 matMRY: This -saf ety evaluation addresses miscotlaneous changes made to the UFSAR as a result of - Deviation - Event Report. DER 90*D429. These changes include the feMowing:-
- 1. _UFSAR section 3.3.2.3.3 - The height of the water covering the top of the fuel storage racks in the spent fuel storage poot (SFP) was revised from 25 ft. to about'22 ft. 6 in. This revision results in about 10 ft, of water above the spent f ust assembtles fonowing a design basis tornado.
- 2. UFSAR section 9.1.4.2.1 = This section was revised to eliminate the reference to the 11ft height restriction of 4 inches and the 10 inch curb to be cleared while troversing the floor with the spent fuel shipping cask.
- 3. UFSAR section 12.1.2.2.2 - The height of the water above the actual f ust portion of the fuel ensembly white in the transf er canat was revised from 7 ft., 9 in, to approximately 7 ft. ,
4 UFSAR section 0.16.7.4 - This section was revised to adopt the results of the f ust handling accident evaluateo in Generat Etoctric Standard Application for Fuel Retoed (CESTAR !!). The change in SFP water depth does not change the function of.the SFP, storage racks. or fuel post cooling system and does not affect any other plant systems. Generat Electric letter TDEC-DE-137 evaluated _ the consequences of changing the depth of water in the SFP and concluded'that it did not affect the consequences- of -the design basis tornado. If such an event occurred, suf ficient water would remain in the SFP to adequately protect and cool the spent fust.. The 22 ft. 6 in, depth is greater than the 22 ft. minimum depth required by Technicet Spectficatton 3/4.9.9. l
. . . _ . __ - . . . _ . _ _ __._.._.___.a._..- - . _ _ _ _ _ ~ . _ . _ _ . . . . , . . ~ . _ . _ _
*, l 4
SAFETY EVALUATIONS PROCEDURES. TESTS, AND EXPERIMENTS Page 18 Safety Evaluation No. 90-0110 (continued): Etiminatim of the spent f ust cask load height restrictions is acceptable in that the restrictions are not the basis f or any accic'ent- analysis. Compliance with NUREO-0612 and the use of single failure proof rigging ensures that a toed drop accident will not occur. The effect of the increase in radiation dose to personnet due to the reduction in the height of the water above the fuel portion of the f ust bundle white in transit through the canat has been evaluated and found to be acceptable and in
-keeping with Fermi's commitment to ALARA. The dose. increase wilt not restrict access to a vital area or otherwise impede actions to mitigate the consequences of an accident.
OESTAR II - evaluated the f ust handling accident f or en assumed drop height _ of 32.95 ft.- The assumed drop height of 32.95 f t. bounds Form 1's actual drop height of approximately. 31 f t. and provides greater operating margin- without sifocting the radiological consequences of the postulated fust handling , accident. The results of CESTAP !! also indicate that 104 fuel rode wilt be ' damaged. This is - bounded by the number of damaged f uel rods - (124) _used to evaluate the radiological consequences of a fust hendting accident.
c.- . - . ., ~ . . - _ - - - - - ~ _ . - . - . - _ - - _ - - - ~ ~ . - ~ .
-4
~ *s e '
+
) SAFETY EVALUATIONS : PROCEDURES. TESTS, AND EXPERIMENTS f Page it .!
?
SAFETY EVALAMTION tlaMARY Safety Evaluation No 90-0122 UFSAR Revision No. 4 Ref erence Doctament UFSAR 3ectioc(s) 9.2 Tabte(s) e Figure Caenge -( 1 Yes (X) No
+
F a Title of Changet RHR Reservoir and Cooting Tower Blocide Controt SlamARY This change revises UFSAR section 9.2.5.4 Tests and Inspections, to an ow the use. of the appropriate blocide . control based on.the evalua*lon of the particular biological fouling problem. The _ originnt UFSAR section g.2.6,4 stated that the reservoir and the cooting tower are periodicaMy inspected for algee growth and treated, as required, with a form of soditmi hypochlorite. The revised section states that the reservoir end- cooling. tower are periodiceMy inspected for macro and micro biological fouling and are treated as required. Being able to trea* for the Specific f outing organism wiu decrease the chance that an organine could block heat exchangers or other. equipment and prevent them from perf orming their saf ety related - f unctions. The scenarios in which E biocide treated water could enter the reactor have b9en evaluated. They are (1) the- injection of reservoir water directly. into the reactor and (2) circulating . water entering the condensate water- system via a condenser; tube leak consequent with decanting the AllR reservoir to the circulating water pond. In the first scenario.-~the blocide concentration would.have U ttle or no - 7 impact . on the ' reactor vesset ce internals due to the high . concentration of impurities which stready exist in the reservoir water. In the second scenario, decenting of the RHR reservoir' is administratively controMed to ensure that the blocido concentration in the circulating water is toss than 1 ppm' or the NPDES . ef f tuent limit _ ( whichever is smatter). _ Since this limit is less than the normat 200 250 ppm dissolved solide concentration of the circulating water, the impurities contributed by the blocide wi n have a negligible impact.
................................................................................ g e
I 0,3-_, . ..+;,,,.s-..4-.--,, . -.~ - . , , , _,,,,.,,_,_,m,-,,r., . , ,_,.,m._..,,m,_ . . _ , , , , . . _ . . -
*s
*, E ;
,e I SAFETY EVALUATIONS PROCEDURES, TEBTS, AND ExptR!MENTS page 20 i
r SAFETY EVALLIATION alteFMY Safety Evaluation Not 90-0127 UFSAR Revision W. N/A Referonce Doctament: 22.000.05 Section(e) Tabte(s) Figure Change ( 3 Yes (X) No Title of Changes - Attornate Means of Measuring Loop h neactor Recirculation Pump Suction Temperature StaAMRY . This procedure change attows the measurement of Loop A Reactor Recirculation Pump suction temperature by using the resistance temperature detector (RTD)'RTD B31-N023A in place of thermocouple B31NO36A. The thermocouple normany f eeds temperature recorder B31 R650 and is inoperable due to an open circuit. The RTD is normany used f or input into the process computer. However, the computer indication does not come on scale untit a reactor temperature of 420 degrees F is reached. .This procedure change anows the RTD toads to be removed from the computer. and connected to a digital inutt ime t e r (OhN) for temperature measurement below 420 degrees F. The RTO toads win be reconnected to the computer when the reactoe temperature is above 420 degrees. The RTO roads the same _ temperature as the thermocouple as they are both in the same thermowou. - There are no seismic concerns as the OhN la sman and located on the f toor and the 06N toads are lugged to the relay panet. The temperature recorder being replaced by the DhN is not required to operate during a seismic event. This temperature readout does not have any control function.'if this readout is lost, plant evolutions involving recirculation temperature wiu be stopped until the readout is restored. I f
, 1 4
SAFETY EVALUAT!DNS PROCEDURES. TESTS, AND EXPERlWENTS j Page 21-SAFETY EVALUATION 9 tat WtY-Safety Evalvotion Not 90-0136 UFSAR Revision No. 4 Reference Document: DC-5241 Section(s) SA.4 Tablut e) f Figure Change i 1 Yes [X) No t Title of Changet Drywen Cap Region Shutdown Analysis and Description Addition to the UFSAR Fire Hazards Analysis
- 9taAMRY:
This safety evaluation documents the shutdown analysis performed in DC-6241 and
- justifies revising section-9A.4.1.11 of the UFSAR to include a description of the drywou gap area.
Design catcutation DC-5241 has anatyred the circuits that penetrate the drywen 1 used for Appendix R thutdown in a fire scenario. This calculation shows that f or these circuits Formi 2 meets the . fire = protection requirements of Appendix R, section !!!.0 for cables in non-inerted containments in that:
- 1. - -The cables that penetrate the containment do not contain . redundant trains of hot shutdown eruipment (the worse condition is a possible effect on one train of hot shutdown equipment).
-2'-
. The cables associated with cold shutdown equipment (valves in this evaluation) can be operated within 72 hours.
- 3. . Associated- circuite that could spurioun ty ' operate valves or tafety retlef valves do not prevent hot or cold shutdown. In addition, there' is no spurious vatvo operation that ~ can create a high/ tow . pressure interface'and overpressure low pressure piping.
A - description of the ' drywen ~ gap area has been added t o :.t he text of ' UFSAR 9A.4.1.11 f ollowing the criteria of UFSAR 9A.2.1 This criteria includes a description of the area, identification of safe shutdown circuits, inventory of combus tibtes, _-- exis ting fire protection equipment, and a conclusion on the impact on safe shutdown capability. There_is no equipment important to safety tocated within the gap other than the penetrations passing through. The results of DC-6241 show that in the event of a-fire in the gap - area, reactor shutdown is achieved in accordance with Appendix R 111.0 guldetines. y g,,- 9y%-rs.-q., +imm+w+~7 +v.- yv - -,,+m,+y p., y, gwy<,--,
1 , l ., f SAFETY EVALUATIONS PROCEDURES, TESTO. AND EXPERIMENTS Page 22 SNTTY EVALUATION 1AM#JtY Safety Evaluation No 90-0141 UFSAR Revision No. 4 Reference Document Sectionte) J5.18 Tabte(s) Figure Change [ ] Yes I X 1 Nt-Title of Change: Single Loop Operation with Foodwater Heaters Out of Service (FWHOS) and/or Moisture Separator Roheaters Out of Service (MSROS) SlM WlY: This safety evaluation evaluates the consequences of having feedwater heaters out of service and/or moisture separator reheaters out of service concurrent with single loop operation. Single Loop Operation with Feeduator Heatore Out of Service Transient and accident analyses were performed for FWHOS during two toop operation. Three abnormat transients were evaluated in data 11. These include: Turbine / generator trip with bypass failure Feedwater flow controtter fatture 100 degree F toss et feedwater heating For the above transients, the two toop FWHOS analysis was performed at the most timiting conditions of 102% power and 100% flow. Since operation in SLO is restricted to 70% power and 75% rated pump speed, these transients are bounded by the existing analyses. The LOCA analysis for two loop FWHOS was evaluated at 104.2% power and 100% flow. The results of this analysis show that reduced final feedwater temperature causes delay in core uncovery times due to the increased system mass and the decreased break flow rate. As a result, the calculated peak cladding temperatures are reduced for FWHOS. Therefore, the LOCA analysis for SLO with normat feedwater temperature is considered to be more limiting than SLO with FWHOS.
_ . _ m _ _ ._.. _.m _ - - _ - - - . . _ . ... -._ - _ _ -. _ . . - . _ . f.
%D c'
SAFETY.EVALUAT1QNS PROCEDURES, T(618, AND EXPERIMENTS page 23 Safety Evatution No. 90-0141 (continued): -l The containement response to pressure and temperature and to dynamio loads is anatyred for two toop operation. The above restrictions on power and pump j speed imposed by RLO ensure that the containment response with SLO is bounded ! by the two toop enetysle. Providing that the existing administrative 1.imi t s for SLO are met, fust cladding and vesset internet integrity is maintained and . the reliability of equipment is not reduced. Additiona M y, no safety limits have been changed. l 1 i Singte Loop Operation with Moisture Separator Reheaters out of Service The cycle 2 retoed analysis includes transients f or MSROS and is described in appendix B of the UFSAR. In order to provide consistency between the cycle 2 retoed analysis end the SLO analysis a review of SLO with MSRSS was performed. The f oMoning cases have been reviewed: Feedwatee contro Mer fatture with MSROS Feodwater contro uer fatture without bypass and MSROS Turbino/ generator trip with MSROS
- Turbine / generator trip without bypass and MSROS-The f eedwater controuer f al(ure with MSROS and turbine / generator trip with MSROS *ransients are bounded by the SLO analysis since the moisture secerator reheater steam flow capacity is toss than the capacity of the bypass system.
The foodwater controner feiture without bypass and MSROS and the I turbsne/ generator trip without bypass and MSROS transients were anatyred at the most timiting power / flow conditions and are described in appendik A of the UFt.AR. Operation in SLO is within the normat power /ftow region and, therefore, is bounded by the two toop reload analysis. , 1 I The SLO and.two toop retoed analyses insure fuel cladding and vesset integrity 4 are maintained white in SLO with MSROS. r
-meeeeeeeeeeeewommeeeeeeeeeeeeeeeemmemememeweeeeeeeeeeeeeeeeeeeeeeeeeeegewe,eoem ,.
- U '
3 y -
.e SAFETY EVALUATIONS PROCEDURf6, TESTS. ANO EXPERIMENTS Page 24 y
J SAFETY EVALUATION SLRAWtY j l Safety Evaluation Not 90-0182 UFSAR Revision No. 4 ! Reference Document UrSAR Section(e) 9.2 febte(s) l Figure Change ( _1 Yee (X) No I Titte of Change: Turbine Building . Closed Cooting Water System (TBCCW) Corroaton inhibiter SLRAWtY: The purpose of this UFSAR change is to remove the names of specific chemicab which - can __ be used f or corrosion control in .the TBCCW system. The origine- . UFSAR safety ovaluation for the TBCCW system specified sodium hydroxide and ! sodium nitrite as the corrosion inhibitors. Thev revised safety evaluation.uses the words " corrosion inhibitors are added for pH and orygen controt". This is ! L the same wording used 'in the UFSAR saf ety evaluation 'f or the Reactor Buitding Closed Cooting Water (RBCCW) . and -. Emergency Equipment Cooting Water (EECW) systems._ By aMowing 'a choice of corrosion inhibitor options, system and component rettabitity is enhanced by anowing system corrosion ' to be minimized -_ using
- state of the art - me thoda. - Since- the new wording te vgebetim to the existing -
; wording : in the saf ety'_ evaluation f or the _ RWCC7f and EECW sys tems, this change ,
does not atter any evaluations in: the UFSAR.' - There . are no new additionet .. burdens on ' the radweste . system due to the addition ef. any corrosion inhibitor -
- in the TBCCW system because of the operationat f toxibility of the radweste l system and because the existing program to minimize t.ny system tenkage w1M be-maintained.
L
% p qrg g u-ey wih m W yrw'ewi- Fw'-' y 5
- Net' y'tq'www- p- yM yt paag me 4-ue--y>a/'e i T P + TJ'e-SMc*w*1- +$87 'iM~d4t$e d
E'FTTf P N** *W7U4-"NL'- -=- P 4D ew- a*bA=
- . _ - ~._
I SAFETY EVALUATIONG PROCEDURES, TESTS, AND EXPERIMENT $ Page 26 SAFETY EVALUATION StaAMRY I
- Safety Evaluation No 90-0158 UFSAR Revision No. 4 l l
Ref erence Doctament UFIMM Secticin(s) 12.3 Table (s) _ , j Figure Change t 3 Yes IX) No Title of Changes Radiation Program Editorist changes SURAMRY: This revision corrects the radiation protection program as descr.ibed in the secticn 12.3 of the UFSAR. The corrections include: Removing reference to protective c\othing and respiratney equipment storage cabinets located at the entrance to the RCA, Protective clothing is stored throughout the plant, near likely locations of use. Respiratory protective equipment is stored in a designated storage and issue area on the second floor of the turbine building.- Removing reference to 'etther checkpoint" in reference to contamination detection and subsequent transit to ti.e personnel decontamination area. In practice there in only one c heckpoin t , with additionat checkpoints established, as needed, during Outages or other periods of high maintenance activity. Correcting the time intervat required for training Edison personnet in radiation protection procedures and techniques that are applicab\e to their a job function. The UFSAR originnuy stated a two year interval In practice such training und testing is completed every year. These changes do not af f act the opeestion of the plant and do not represent any fundamental changes in phttosophy or procedures. The corrections meet the opplicable- design' criteria specified in section 12.3.1.2. of the UFSAR. Protective clothing and respiratory equipment are stin avail,able, the integrity of the RCA 16 maintained, and more frequent training results in better trained personnet. k
-- +- - - - - - -, . . , , ~ . - . _ . . .- , . . . - - , - , , ------n..- , ,n. , ~ ~ - . , , , , , ,, ,,,=,nne.-,,,---,- w
. . - - . . , . _ . ~ . ..~ - ..- --.-.-.. . -
s 4 .e'
,a SAFETY EVALUATIONS PROCEDURES, TESTS,-AND LXPER!iAENTS Page 20 SAFETY EVALUAf!ON 1RAAMRY ,
k Gafety Evaluation Not _90 0159 UF&AR Revision No. 4 Section(s) Reference Document: UFSAR SA.11 13.2 Table (s) Figure Change ( ) Yes IX) No Title of Changes Fire Protection Program Personnet Organization - Changes and Firo Ori u Schedute Ctarification
#8 sh%RY This change revises section 9A.1.3.3.d.1 of the UFSAR to state that the Fire
-Protection $pecialist reports to Ptant Befety. This section originalty stated that the ' Fire Protection -Specialist was within Nuotone Production. In addition. this change revises section 13.2.4.2.1.3 of'the UFSAR to state that drius with the of f site fire department shau be perf ormed once per calendar year. This section previously stated that these drius would be performed at tenst annuaMy.
These changes are: administrative in nature since . the requirements, plant configuration, curveinance. - compensatory action and program imptomentation remain unchanged. The fire protection spectatist responsibilities wlu.not be attered by - this change. The requirement to perf orm a drin with the of f site fire department once.sach year meets the original intent of the fire protection
' program.
l l i .' i
. - . . ~ -, . . . . . , . -- - _ - . _ _ , _ _ _ _ . . . . , ...m.-.. -,.m . , . . . . . , _ . - - , . - . . _ , . - . . - . . . . .
a CAFETY EVALUATION $ PROCEDURES, TESTS, AND EXPERIMENTS Page 27 EAFETY EVALUATION tAA&MRY Safety Evaluation No 90-0161 UFSAR Revision No. 4 Reforonce Document: DER 89-0769 Sectionte) Table (6) 1.6-2 Figure Change tX1 Yes ( ) Ho Title of Change: UFSAR Figur- 'ipdate and Conversion to Detroit Edison Drawings Bl84%RY: Thie safety evaluation justifies the updating and conversion to Detroit Edison drawings of nineteen UFSAR figures. These figures are simplified flow diegrams for the Residual Heat Removot, Core Spray, High Pressure Coolant Injection, and Peactor Core Injection Cooling systems. Deviation Event Report 89-0769 was written because th6se figures were out of date. As part of the corrective action, the figures were updated to reflect the as-built condition of the plant based on exleting approved Detroit Edison piping and instrument drawings, Conversion to Detroit Edison drawings ensures that the UFLAR figures will reflect the as-built status of the plant af ter f uture plant modifications are incorporated. Since these chanDes involve drawing updates and redesignations to ensure that the UFSAR figures reflect the existing as-built status of the plant, they do not make any physical chanDos to the phnt and do not affect the function or operation of plant equipment.
.f
/ fM ETY EVALUATIONS PRDCEDURES, TESTS, AND EXPERIMENTS Page 28 SAFETY EVALLATION SLA449tY Safety Evaluation No 91 0003 UFSAR Revision No. 4 Reference Documentt , UFSAR _ Sectlon(e) 7.3 Table (s) _ Figure Change ( ) Yes IX) No Title of Changet Primary Containerent Isolation Logic and Containment Pressure Transmitter Description Changes and Corrections SO4AVtY This change revises the description of the primary containment isolatic9 logic contained in UFSAR section 7.3.2.2.4 and corrects the number of transmitters monitoring primary containment pressure described in UFSAR section 7.3.2.2.0.4. This change is the result of an independent Safety Engineering Onoup (ISEO) technical review of the high drywell pressure signal instrumentation. The isolation logic debeription was revised to provide a logic description f or inboard and outboard contaireent isolation valves other than main steam isolation valves, The number of transmitters monitoring primary containment pressure was changed from 0 to 4 This change was made to reflect the actual plant configuration. This change has no impact on the pipe break analysis described in UFSAR sections 6.2 S.3. and 1646.6. The revisions are in agreement with Technicat Specificatinns 3/4.3.2 and 3/4.6.3 concerning primary containment isolation actuation instrumentation and isolation valves.
2 t E SAFETY EVALUATIONS PROCEDURES. TESTS, AND EXPERIMENTS Page 29 the following Technical Specification Amendments were incorporated into Revision 4 of the UFSAR. The NRC safety evaluation (which la based on the Detroit Edison evaluation supporting the change) that iccompanies each amendment provides the basis and justification f or the UFSAR revision. T S. Amendment UFSAR Section/ Table 41 Tebte 7.6-2 60 1.4 63 16.0 16.10 64 13.1 13.4 60 Table 7.6-2 Table 7.3-9 62 9A.1 END OF BAFETY EVALUATION SlamARY REN)RT
------___.______.----_----_.,--_--_--_-.-_-__-.--.---_----._--_.---.--_.----_.,.-_--___----_.-__.-__.-__-----..-._-----____--._-_-_--_-_.---__.-------___J}}