NRC-90-0041, Safety Evaluation Re as-built Notices
| ML20012C309 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/31/1989 |
| From: | DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20012C302 | List: |
| References | |
| CON-NRC-90-0041, CON-NRC-90-41 NUDOCS 9003210014 | |
| Download: ML20012C309 (103) | |
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.t Enclosure to NRC 90-0041 u
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FERMI 2 4
SAFETY EVALUATION
SUMMARY
REPORT 1989 I
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Docket No. 50-341 7
License No. NPF 43
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1 FERMI 2 SAFETY EVALUATION
SUMMARY
REPORT i
1989 1
AS-BUILT NOTICES g
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s Safety Evaluations ABN's Page 2 i
AS-BUILT NOTICES (ABN'S)
Safety Evaluation No.: 87-033g, Rev. 1 Figure Change 2mplementation Doc e nt No.: ABN 6775-1 System No.: W2300 Title of Change: Circulating noter Drawing Revisions Summary:
The f ollowing drawing changes were made to the system P&ID: 1) a section i
of the drawing was expanded / relocated to illustrate all four chlorinators so the correct piping configuration can be shown 2) filters were added to the inlet lines of the f our chlorinators, 3) a missing valve was added to one of the chlorinators, 4) en incorrect valve designator was corrected.
-and 5) en instrument and controls reference drawing number was revised.
Safety Evaluation Summary:
The changes to the system P&ID enhance the description of the chlorination system. By expanding the portion of the drawing to show all four chlorinators es they appear in the field, the drawing becomes more detailed and it is now possible to trace the flow through the chlorinators without any guess-work required. TPm other drawing changes correct minor discrepancies between the P&ID and actual field conditions or current drawing designation. The chlorination systems' primary function is to i
enhance operation of the circulating water system, it is not required f or I
the saf e shutdown of the plant. Since the changes are consistent with actual field condition as well as the original system design, there is no af f ect on the saf e operation of the system.
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Safety Evaluations
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Safety Evaluation No.: 88-0124 Implementation Document No.: ABN 8668-1 System No.: N2102 Title of Change Reactor Food Pump Design Suction Pressure Change Summary:
This as-built notice addet antes to the reactor f eed pump (RFP) vendor manual and purchase specification stating that the pump suction was acceptable considering suction design pressures as high as 820 pounds per equare inch, gauge (psig).
Safety Evaluation Sunenary:
This as-built notice ef fectively increased the design pressure of the reactor feed pump (RFP) suction from a design value of 395 pounds por square inch,,guage (psig) to a design value of 820 psig. A fini'te element analysis of the critical reactor f eed pump suction nozzle was conducted by Hopper and Associates, considering internal pressure and attached piping leads. Pump casing material was tested in order to determine material properties as well as maximum allowable stresses. The analysis concluded that the RFp's are safe for operation at temperature with internal pressures as high as 820 peig concurrent with computed piping reaction loads.- As discussed, the reactor' food pump suction satisfies the applicable design requirements for saevice at 820 psig. Further, the 820 psig design pressure is adequate for pump suction design purposes, considering pump operating conditions and equipment malf unctions.
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Safety Evaluations ABN's' Page 4 Safety Evaluation No.1 69-0029 Figure Change Implementation Document No.: ABN 6689-1 System No.: N3000 Title of Change: Turbine Generator System Drawing Revisions.
Suunary:
The following revisions were made
- 1) the generator gas system instrument flow diagram and f unctional operating sketch were revised to delete piping from the hydrogen distribution and carbon dioxide headers to gas coctors number 1 to 4; 2) the general service wette disgram was revised to remove the generator gas (hydrogen) system instrument loop; 3) CECO was revised to change operating temperature range of the generator gas (hydrogen) system thermocouples f rom 86-122 F to 125-140 F and 100-105 F for hot and cold gas respectively 4) drawing cross references were added to general service watee. generator gas and turbine electrical / instrumental system drawings.
Saf ety Evaluation Sunnary:
The changes made by this ABN removed the generator gas (hydrogen) system instrument loop f rom the general service water system diagram. This loop was atroady shown on generator ges system instrument flow diagram. This change was completed to maintain each system's instrument loop on their respective P&lD and eliminate the duplication. The general service water system and generator gas system are non-nuclear non-seismic systems designed to remove various heat loads f rom the plant and main generator, respectively during normal operation.. The general service water system and generator gas system are not safety related systems and do not have any adverse ef f ects on the saf a shutdown of the plant.
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~1 Safety Evaluations ABN's page 5
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.l Safety Evatuation No.: 89-0224 Implementation Doounent No.: ASN 10859-1 i
System No.:.N3011 Title of Change Removal of 5th Stage Slades on Low Pressure Turbine l:
Suunary:
A11 5th Stage blading (both rotary and fixed diaphragm) were removed from l'
the three Low Pressure turbines due to cracking and f ailure of the rotating blades. This problem.was discovered during the RF-01 Turbine Inspection that was required as a result of turbine vibration problems..Trw turbine manuf acturer has recommended operating with the entire 5th stage removed untiW replacement blading is redesigned and manufactured. Removal of the 5th stage will change the perf ormance of FWHTRS #3 and #4, because the extraction steam pressure to both of these heaters will be reduced.
t' Safety Evaluation Summary:
'This change was evaluated against the postulated f ailures within the power conversion system described in UFSAR Section 10.1.
The evaluation considered breaks in FW system piping and the air-injector line, the turbine missile analysis, and introduction of contamination into the RPV.
i The revised operating parameters of the condensate, feedwater, extraction steam and heater drains systems are within their respective design limits and the turbine manuf acturer has stated that the low pressure turbine i
blading and diaphragm stress levels are within permissibts design values.
The alight reduction in turbine passive bypass steam flow was evaluated against the various transient analyses in Chapter 15 sf the UFSAR. The reduction in paselve bypass capacity of the reheater and the reduction in final f eedwater temperature are bounded by the UFSAR Chapter 15 Analysis and theref ore the margin of safety as defined in T.S. Bases - 3/4.2.3 (Minimum Critical Power Ratio) and 3/4.7.9 (Main Tur.<ine Sypass Systems and Moisture Separator Rehester) has not been reduced.
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Safety Evaluation No.: 90-0003 UFSAR Sections SA.4 &
9.5 Figure Changes Implementation Document No.: ABN 10884-1 1
i System No.: P80 and Various Title of Change Fire Protection Drawing Changes
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Sunenary:
This os-buit,t notice has been written to dif f erentiate between Technical Specification and non Technical Specification related fire separation barriers on fire protection evaluation drawings and corresponding UFSAR Figures. The associated safety evaluation also addresses the correction of miscellaneous typing errors. text duplication and valve identification numbers in the UFSAR.
Saf ety Evaluation Sununary:
This as-built notice does not involve a physical plant modification. The changes to fire protection evaluation drawings and corresponding UFSAR Figures and text changes do not degrade the Fire Protection Program because they are made to correctly reflect the as-built plant configuration. These I
r changes do not af f ect the accident analysis or the operation / function of any safety-related equipment, i
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DESIGN CHANGE StaARRY l
Safety Evaluation No.: 90-0009 Figure Change implementation Document No.: ABN 6517-1 System No.: N7100 Title of Change: Circulating Water Pump Cooling Water Drawing Changes Saunary:
This ASN revised the circulating water (CW) and General Service Water (GSW) system P&lD and FOS drawings to reflect the as-built condition of the cooling water supply to the CW system pumps. The change reflects removat of c p lone separators that were used to provide additional cleaning to the GSW which is used to cool the CW pumps' lube oil and lubracate the CW pump bearings. The cyclone separators were replaced with a new 6" supply header, a 2" line to each CW pump and associated valving for system isolation.
Safety Eval.uation Summary:
The cyclone separators were removed because the eman 1/2" inlet supply lines became partiauy clogged by washed away loose scale deposits, which reduced cooling and lubricating flow to the CW pumps. This change win assure a continuous flow of cooling and tubricating water to the CW pumps resulting in more reliable pump operation.
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i Safety Evaluations ABN's Pepe 8 Safety Evaluation No.:
89-0172, 89-0173, 89-0198, 89-0199, 89-0200, 89-0202, 89-0203, 89-0205, 89-0206, 89-0207, 1
89-0209, 90-0004, 90-0011, 90-0024 Implementation Document No.
Seo Below System No.: Plant Wide j
Title of Change Locked Valve Program Changes 1
Sumnery:
In response to Notice of Violation (50-341/88021-01), 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 1 through 28 were issued to document the results. Design Calculation 4959 Volume 1' represents the Mechanical portion of the design calc, and Volumes 2 through 28 represent the I&C portion of the design calc. The design J
calculations identify Mechanical and I&C valves that are required to be
-I locked. Fonowing issuance of Design Calculation 4959 Volumes 1 through 28, changes were approved and implemented to summarise the results of the
__,f Design Calculation and valves were subsequently locked in their in-service pesition. During the review, discrepancies were identified which required corrections to plant procedures and to system drawings to make them consistent.
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Safety Evaluation Sununary:
l 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 system win be mispositioned. Adding locking devices to velves win secure the valves in their in-service position. This gives added assurance that systems and equipment important to safety are operable. The addition of looking devices does not affect the function of the valves (or other equipment) other than adding a physical barrier to help prevent mispositioning of valves. The weight of the locking devices is emau compared to the weight of the valve and wi n have an insignificant offact on seismic considerations.
Y Safety Fvelvetions ABN's Page 9 Safety Evaluation No.: 90-0033 UFSAR Bection 12.1.4.3.g UFEAR Figure Changes Implementation Document No.: ABN 7072-1 System No.: D21 Title of Change Area Radiation Monitoring System Summary:
This as-built notice removes the Area Radiation Monitor's (ARMS) alarm range and setpoint inf ormation f rom UFSAR Figures 12.1-12 and 12.1-13.
This inf ormation is contained in and controlled by the Central Component Data Base (CECO). This as-built notice also provides the justification f or changing the setpoint f or ARM D21-K730 to prevent nuisance alarms.
Safety Evaluation Samary:
The ARM alarm ranges and setpoint inf ormation in the UFSAR is also contained and controtted by CECO. The CECO system is administrative 1y controtted and administrative procedures direct interested individuals to find this type of information in the CECO system.
F Sofety Evaluations ABN'a Page 10
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k The following as-built notices were issued to correct drawings and UFSAR figures under the locked valve program.
P ABN 10900-1 ABN 10915-1 ABN 10901-1 ABN 10917-1 ABN 10903-1 ABN 10919-1 ABN 10904-1 ABN 10920-1 ABN 10905-1 ABN 10921-1 ABN 10906-1 ABN 10922-1 ABN 10907-1 ABN 11177-1 ABN 10908-1 ABN 11187-1 ABN 10909-1 ABN 10910-1 ABN 10911-1 ABN 10912-1 ABN 10914-1
Safety Evaluations ABN's Page 11 The following As-Built Notices ( ABNs) and Engineering Design Packages (EDps) 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 as-built plant conditions a summary for each was not prepared..The ABNs or EOPa and their associated saf ety evaluations have been listed for reference.
Safety Evaluation No.:
SEC 1843 Figure Change
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Implementation Document EDP 1843 Safety Evaluation No.:
86-0128, 89-0001 Figure Change Implementation Document ABN 8441-1 Safety Evaluation No.:
88-0240 Figure Change l.
Implementation Document ABN 8434-1 Safety Evaluation No.:
89-0010 Figure Change Implementatiove Document ABN 9165-1 Safety Evaluation No.1 89-0034 Figure Change
- Implementation Document ABN 8205-1 Safety Evaluation No.:
89-0040 Figure Change Implementation Document ABN 5490-1 Safety Evaluation No.:
89-0043 Figure Change Implementation Document ABN 6697-1 l
Safety Evaluation No.
89-0113 Figure Change 2mplementation Document ABN 10120-1 Safety Evaluation No.:
89-0119 Figure Change 2mplementation Document ABN 10529-1 Safety Evaluation No.:
89-0149 Figure Change l
2mplementation Document ABN 10598-1 1
Safety Evaluation No, 89-0174 Figure Chang 6 Implementation Document-ABN 9912-1 Safety Evaluation No.:
89-0222 Figure & 70st Channe Imptomentation Document ABN 10342-1 Safety Evaluation No.:
09-0228 Figues Change l
Implementation Document ABN 10767-1
Safety Evaluations ABN's Page 12 Safety Evaluation No.:
89-0230 Figure Change 2mplementation Document ABN 10001-1 Safety Evaluation No.:
89-0231 Figure Change Implementation Document ABN 10303-1 Safety Evaluation No.:
89-0233 Figure Change Implementation Document ABN 10459-1 Safety Evaluation No.:
89-0234 Figure Change Implementation Document ABN 7196-1 Safety Evaluation No.:
90-0001 Figure Change 2mplementation Document ABN 9553-1 END OF ABN SECTION
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FERMI 2
- SAFETY EVALUATION
SUMMARY
REPORT 1989 ENGINEERING DESIGN PACKAGES
Safety Evaluations EDP's Page 2 ENGINEERING DESIGN PACKAGES (EDP'S)
DESIGN CHANGE SEASMRY Safety Evaluation No.: 86-0137 UFSAR Tables 6.2-2 8.2-13 & S.2-15 Figure Changes implementation Document No.: EDP 4789 System No.: T6000 Title of Change: Removal of Primary Containment Isolation Valves 86sumery:
This EDP removed two air operated primary containment isolation valves, one f rom each of two primary containment pipe penetrations, and one manual valve from one of the pipe penetratiens. The lines were capped, and several pipe supports and one snubber were removed. The lines had been used f or drywell air sampling to H /0 m nitors and were no longer required.
2 2 Safety Evaluation Summary:
The removal of these valves required a Technical Specification change to Table 3.6.3-1, which was issued as Amendment No.16 to license NPF-43 on March 21, 1988.
Safety Evaluations
.EDP's PeDe 3 DES 10N 04ANCE SLASAARY Safety Evaluation No.: 86-0193 Figure Change Imptomentation Dooment No.: EDP 2216 System No.: 011 Title of Change: Addition of In-Line Fitters for Controt Center Emergency Air Radiation Monitors Summary:
This modification adds in-line filters to the Control Center Emergency Air Wknitors D11-P285. D11-P290, 011-P297 and 011-P298. The fitters are being added to prevent crud build-up on detector surf aces and to provide a higher differential pressure drop. This increased pressure drop will decrease the process fluid pressure such that it will be towered into the detectable range of the instatted pressure switch. This pressure switch provides a control room troubts alarm on high and low system pressure.
Safety Evaluation Summary:
The subject monitors are used to detect noble gases and the addition of the filters will prevent particulate matter f rom f ouling them.
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Safety Evaluations EDP's Page 4 DESIGN CHANGE SLBSAARY Safety Evaluation No.: 86-0233 Figure change Implementation Document No.: EDP 3644 System No.: P80, TSO, X80. U80 Title of Change Installation of Pressure Gauges Samary:
This EDP added a pressure gauge to fire protection system risers. These gauges were added to comp 1,y with NFPA 14, which requires pressure gauges to be located at the top of each fire hose standpipe.
Safety Evaluation Summary:
The addition of these gauges provides compliance with the NFPA code and plant design basis. There is no impact on safety reisted systems.
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EDP's Page 5 DESIGN CHANGE SLbbaARY Safety Evaluation No. ' 86-0266 Figure Changes Implementation Doo m nt No,t EDP 6087 System No.: D4100 Title of Change: Temperature Indication of SLC Pump Suciton Summary:
This modification added a temperature sensor and local temperature. indicator to the standby liquid control pump suction piping. The indicator attows local verification of SLC's pump suction piping temperature required by Tech Specs.
Safety Evaluation Summary:
The sensor is a surf ace mounted device, and did not require a pipe penetration.
The sensor and indicator were seismically mounted on the SLC pump suction piping. The addition of this equipment does not af fect the function er operating characteristics of the system, but provides local temperature information only.
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Safety Evaluations EDP's Page S DECION CHANGE SLANARY Safety Evaluation No.: 87-032g UFSAR Section 9.2 Figure Chsnees Imptementation Doctament No.: EDP 7557 System No.: P4300 Title of Change Station Air Compressor Cooling Water Alarm and Temperature Indication Stannary:
inis EDP reptaced existing tocat thermometers that indicated cooling water temperature tesving each station air compressor tube cit cooter with resistence temperature detectors and electronic indicating temperature contreu ers.
AdditionaV.y. tbs cooting water low flow alarm was modified to annunciate on l
high water temperature tesving each station air compressor tube ett cooter.
Safety Evaluation Sununary:
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Frequent no flow steeming of the Turbine Buitding Closed Cooling Water System to the Station Air Compressors was being experienced due to changing plant conditions. This no flow alarm which is based on cooling water differentist pressure across the air comprescor did not annunciate in au cases when flow was nortiau y tost to the compressor tube oil cooters. The alarm scheme was changed to add a temperature sensing unit to the differentist pressure measurement to
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directly reflect the variable of concern., As stated in UFSAR g.2.7.3 turbine auxiliaries housed in the Turbine building are not considered essentist to safe-L reactor shutdown.
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-e Safety Evaluations EDP's Page 7 DESIGN CHANGE SLSNARY Safety Evaluation No.: 88-0007 4/MR SECTIONS LISTED SELOW Implementation Document No.: EDP-6509 System No.: 01120, G1135
. I Title of Change Redweste System Changes Summary:
EDP G539 modifies the flush water supply to the centrifuge by removing pressure control valve FA08 and restriction orifices R0-D001 and R0-D002. This equipment is being removed to increase the flush water flow to the contrifuge *o reduce residual contamination. Changes were also made to the redweste automatic transfer and flush sequences to provide additional pump / tank interlocks to i
f acilitate the completion of 11guld system testing and reduce the probabil,ity for spi n s dUring tank transfers, i
Safety Evaluation Summary:
In the UFSAR, the Radweste System is evaluated for seismicany-induced simultanoous fsilure of au Radweote syatom tanke. The modifications msde by this EDP do not result in the increased probability for or consequences of an accident evaluated in the FSAR.- Moreover. the changes being made to the system are such that they reduce the probability for spius and reduce the probability
'for personnet exposure that cout,d result from those spiu s.
There are no l-Technical Specification sections associated with any portions of the redweste-system associated with this EDP.
The fonowing P&!D revision = resulted f rom this EDP and af f act the listed UFSAR j
Figures pSID UFSAR Figure SM721-2011 10.4-7 SM721-2215 11.2-6 SM721-2222 11.2-7 l-L SM721-4941 11.2-6 I
SM721-4942 11.2-9 i -
SM721-4943 11.2-1D SM721-4944 11.2-11 SM721-4945 11.2-12 SM721-5094 11.2-13 SM721-5122 11.2-14 I-
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i Safety Evaluation No.: 88-0011 1setamentation Doeumont No.: EDP 8106 System No.: T2300. T4900 Title of Change Vacuum treeker Valve Modificatiens i
Summary:
This modification replaced the carbon steel shaf t and wafer with stainless steel components in the primary containment suppression chamber to a reactor building vacuum breakee valve end a nitrogen inerting suppression pool vont valve.
I Safety Evaluation Summary:
The proposed change modifies the materiet used f o' the valve's shaf t and water, but does not a\\ter the function of the valves. Substituting staintese st.4 for carbon stoet for the shaf ts and waf ers enhances re\\tabi\\ity by reducing tio
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potentist for corrosion product buildup. The modified valves are designed and tested to meet the same criteria as the existing vetves. Parts are reptaced on a one-for-one basis, and no additiona\\ parts are being added to the va\\ves, hence no new moJie of valve f ailure it created by the modification.
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i DE82tM CHANGE SLANARY Safety Evaluation he.: 88-0066 rigure Ctence 1mptementation Doo6snont No.: EDP 1611 System No.: P6000 i
Title of Changes Station Air Compressor System Piping Modifications l
Sunenery:
The station air compressor system drain piping and in-line components were modified to improve water / moisture drainage and eliminate corrosion product buildup.
I Safaty Evoluetson 86sunary:
The station ear system is not required f or the saf e shutdown of the plant. This modification was implemented to enhance the existing design. The station air system is a non-saf ety. CA level !! system and the modification does not af f act any safety system.
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Safety Evolustions EDP's I
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Safety Evaluation No.: 88-0095 Fly >ee Change 3aplementation Dooumont No.: EDP 8938 System No.s N2200 I
Title of Change Valve Modifications Swanary:
The purpose of this modification was to instan a blank plate inside the body of a north separator seal tank valve. The valve seat and disc had been samaged beyond repair, and the plate was instaMed to eliminate bypass lookape.
Originauy, this valve was ietstaued in the moisture separator seal tank drain line to the condenter to bypass and protect a tevel contro\\ valve during flushing of the moisture separater reheater.
Safety Evatuation Sunmary:
The moisture separator seal tank s; rain \\ine f unction and operability are not af f acted by the change ar.d furthermore the inst 4Mation of the blank plate does t
not affact the accident or transient analysis. The insteMod plate and specified weld were sized based upon system design conditions. The proposed i
change does not ef f ect any equipment integral to plant normal operation or safe shutdewn. Since the negative effects of the damaged valve Geekogo) are removed overe n system re\\tability and performance are enhanced. The separator seal tank drain function is maintained.
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Sofety Evaluations
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EDP's Page 11 DEllGN CHANGE SLA4WLY Safety Eve \\untion bio.: 84-0146 Figure Changes
!aptamentation Document No.: EDP-0283 System No.: T4100 Addition of Vent pipe to Reactoe tutteing Heating Title of Change Summary:
This modification invo\\ved adding a new 4" vent pipe f rom the condensato storage The tank, which ta \\ocated tank of the reactor bu1\\ ding steam hosting system.
in the reactor bu1\\ ding sub-besoment Division 11 Core Spray corner room, previously was vented with a 2" pipe to the reactor buitding refuo\\ing ficer.
Jt to now vented oute16e the reactee butiding which is part of secondary A\\so the tank's \\oop seal and overfiow pipe were removed and the containment.
openings capped.
Safety Evaluation h ry:
The saf ety evaluation considered ef fects on maintaining secondary containment integrity, the ability of SGTS to perform its functions, and radiciopical reioase pathways 6ven if the vent were to break creating a d opening in Also considered were flooding, structurai integrity of ascendary containment.
the vent, security, and missite protection. The of f-site dose ana\\ysis performed for a design basis accident remains vottd with the implementation of this modification. This modification does not reduce the margin of saf ety as defined in the bases f or any Technical Specification.
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Safety Evaluations EDP's pope 12 a
DESIGN CHANGE SLAMARY Sefety Evatuotion No.
46-0196 F1gure Change UFSAR Section 3.2. Table 4.2-1 S 8.2-2 Imp'tementation Desument No. 3 EDP g248 System No.: $4003 Title of Change: Substetton Name Change Summary:
This modificetten changed the 120KV Enrico Formi Brownstown line to Enrico Formi Swan Creek line due to the insten ation of the Swan Creek substetton. A6 part of the thenge-over, this modification was developed for positten W on the 120KV Due 102 to replace a 400 emp wave trap with a 1600 emp wave trap, replace the first line relaying penet 1 A replace outdated HIM relays with OCX61 A distance rettys and add SA#148 and AR retsys. Due to the revised name of the line, the updated final safety analysis report (UFSAR) w1M be revised to reflect the new name and new Swan Creek substetton.
Safety Evaluetten Summary:
The additionat twen Creek substation, relaying upgrade, and wave trap espacity improvement af f act the 120KV position W and the 120KV trenomission line. AM of the above are part of the 'of f site" power supply f or Formi 2.
Accident anatysis f or Formi 2 essumes a loss of the 'of f ette" supply. The Fermi trownstown tine was not retocated but renamed. There is ne change to the time er towers exiting the plant as shown in the updated finat esfety analysis report (UFSAR). The first time protective retsying philosophy remains the same even though the relays are being reptooed. The Fermi 2 degraded grid storms and undervottage relaying are tocated within the Fermi 2 design and are not affected by the change made to the Formi i reteying. AM of the equipment in this modification to non-safety colated and was made to one of the of f aite 120KV foods to the plant.
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DESIGN CDMNQE SIAS4ARY Safety Evaluation No.: 88-0218 UFSAR Section 3.8. S.2 T.S. Table 6.2-2 and Figures Changes Implementation Doement No.: EDP 4940 l
l System No.: C5100. CB102. T4901 Title of Change Change in Use of a Primary Containment Penetration Susanary:
This EDP relocated the transversing in-core probe (TAP) purge control velve penet inside the drywen. By tocating the purge controt panet inside primary containment it was possible to supply nitropen to the penet from an existing suppty line inside the drywen and eliminate the use of a primary containment penetration. The flange of the penetration was removed and a pipe cap welded in l
Ats place making this penetration a opere. Thus, the general design criteria (ODC) requirements f or this penetration no tonpor apply and this penetration l
becomes a 10CFR60 Appendix J Type A test penetration.
Safety Evaluation Summary:
Eliminating the use of the containment penetration win reduce the probabnity of release of radioactivity due to the elimination of active components associated with the penetration. The penetration was provided with a socket
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welded cap. qualified and tested to appropriate ASME standards. Further, the nitropen line for T!P purge was provided with a restrictive orifice. This is designed such that a ruoture of the TIP purge line win not degrade the perf ormance of the other primary sentainment pneumatic supp\\y (PCPS) functions supplied f rom the main PCPS f eed inside sentainment. The design also assures that even when the main PCPS feeder is at minimum design supply preesure. At wi n provide sufficient flow to s u indemore for TAP purge.
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l Safety Eve \\uations EDP's
,e.e,e DES!ON CHANGE SLSSWtY j
I Safety EveLuetten No.: 88-0223 UFSAR Section 9.3 Figure Change Zeptementation Dooumont No.: EDP 5323 System No.: P5002 Titto of Change: Addition of Interruptib\\e Air Supply Dryer Unit r
Sasumery:
The purpose of this modification was to procure and instatt a second air dryer / fitter unit in the non-saf ety related interruptible air suppty (!AS) system.
Safety Evaluation Samunary:
With the imptomentation of this modification. the interruptible air supp\\y (IAS) system will have 100% backup plus bypass capabilities. The capabitity of the oystem to be supplied with clean dryer air was improved by the addition of a
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second 100% cepecity unit. Sesed on these facts, the risk of toes of the IAS l
eystem has been decrossed.
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EDP's Page 16 l
t DES!(34 CDMNGE SLhhERY l-t Safety Evakastion No.t 48-0021 Figure Change 1mphaentation Deeunent No.: EDP 9040 Systen No.
C1100. P1100 Title of Change Condensate Storage Tank Check Valve Addition Susunary:
This modification added a six inch stainless steet swing check valve in the suppty header f rom the condensate storage tank (CST) to the control rod drive (CRD) pump suction.
Safety Evehantion Summary:
The postulated lose of the condensate supply to the control rod drive (CRD) pumps, core spray pumps, or standby feedwater pumps has been evaluated due to the potential loss of the supply header, which is non safety related. - The supply header is not required for the sofo shutdown of the plant. The addition i
of the check valve assembly does not af f eet oef e plant shutdown. This modification was intended to eliminate an excessive alarm occurrence and to improve the CRD system's performance. Adding the check valve does not increase the probability of postulated f ailure of the supply header in which it is instaMed.
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Safety Evaluations EDP's Page 16 i
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DESIGN CHANGE SLANARY l
Safety Evaluation No.: g9-0036 Figure Change laptementation Deoumont No.: EDP g969 System No.: E1160 Title of Change Residual Heat Removat Valve Modifications r
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l Modification was implemented to revise the control scheme of the residuet heat s
removat (RHR) recirculation outboard isolation valves to enable throttling the valves in the manuat opening mode. The previous controt scheme allowed l
throttling in the manual close mode only. The automatic control (accident condition), was not altered.
Safety Evetuation a6sumery:
In the event of an accident. loss of coolant accident (LOCA) or high energy line l
break (HELS), the low pressure coolant injection (LPCI) toop selection logic is automatically initiated on high drywell pressure or low reactor water level.
This logic ensures LPCI injection into en unbroken recirculation loop by energising the open motor control circuit for the residuet heat removal (RHR) recirculation inboard and outboard isolation valves associated with the unbroken recirculation loop. The loop selection logic also energises the close motor controt circuit f or the intoard and outboard valves associated with the broken recirculation loop. This action peevente RHR pump flow f rom being diverted to a broken toop and defeating the purposa of flooding the core.
Since the inboard end outboard valves are required to theettle LPCI flow during sortain eenditione, this modification reinstated the throttling capability for those valves in the opening direction. This enhanees the sentrol scheme of throttling in the steeing direction only and returne the valves to their originat design eenfiguration. It also maintains the intent of previous modiftsations by maintaining the limit open cutout to peevent backsoating.
Safety Evaluations EDP's page 17 DE81tpd CHAM 1E SLthERY Safety Evatuation No.: Sg-0042. 89-00$8 Figure Change Implementation Document No.: EDP $331. EDP 10117 System No.: P4400 Titto of Changes Emerger.cy Equipment Cooling water Temperature Recorder Replacement Susunary:
These modifications replaced the damaged temperature recorders on the outlet lines of the emergency equipment cooling water (EECW) heet exchangers. The damage recorders (Sailey Mode SR) were replaced with Leeds & Northrup (L&N)
Model 134 The L&N recorder is a similar component replacement, accepts the same input signal has the same range (0-200 F), and is physically interchangeable (with an adapter plate). No system function or operation was changed.
Sefety Evaluetion Summary:
These modifications replaced obsolete recorders which are no longer manuf actured and many apare parts are no longer available. The original escorder model required an isolator to prevent any possible offacts of multiple grounding of inputs which would affact the measurement process. The new recorder model eentaine an internal power supply which eliminates the possibility of multiple grounding and eliminated the auto-isolator from the circuit. The teolator is shown en the emergency equipment elooed cooling water (EECW) system diagram.
Replacement of the temperature recorder with a simitar sempenent and removal of the isolatoe did not change any system function er operation.
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Safety Evaluation No.: 09-DD49 UFSAR Section S.3 4 7.3 Tables 7.3-2 & 15.6.4-1 Figure Changes Implementation Doonsnent No.: EDP 1022 l
System No.: 32104 Title of Change Automatic Depressurisation System Logic Modifications I
86sunary l
The automatic depressurination system (ADS) logic was modified to show the l
system to automaticau y actuate after reactor pressure vessel (RPV) tow water level 1 has existed f or 7 minutes. The previous system required both RPV level 1 and high drywe n pressure for automatic actuation. This modification win also provide manual inhibit switches and alarms for procedureu y directed operator intervention into ADS operation.
Safaty Evaluetten Sununary:
The changes enhance the system's capabilities because it relieves the operations statf of having to manuan y initiate the automatic depressurination system (ADS) during postulated events where reactor water tevel remains low but drywe n pressure does not increase. Any inadvertent operation of the manuel inhibit l
ewitch win be identified by actuation of a white indicating tight and alarm annunciation such that corrective actions can be taken. Furthermore, the manual inhibit switch was keytocked to prevent unauthorimod use. Neither the overpressure relief function, the manual ADS initiation, safety relief valve, nor individual (SRV) contret wi n be affacted by operation of the manual inhibit switch or drywen pressure-high bypass timer modifloatione. Likewise, the modification does not affact the high drywe n pressure soincident with low water level initiation sequence for pipe besaks inside containment. The proposed changes are bounded by existing updated final safety analysis report (UPSAR) analysis, inoeoase the system's sapabilities by automaticany actuating ADS for a geester number of postulated events and do not eliminate por diminish the existing system capabluttes. NRC issued a safety evaluation in support of License Amendment No. 35 which addressed this modification.
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Safety Evaluations EDP's Pepo 19 DES 10N CHANGE Sta8MRY l
Safety Evaluation No.: 89-0050 UFSAR Sections 4.8 8 7.4 and Figure Changes Imptesmontation Doceent No.: EDP 1021 System No.: C4100 l
Title of Change Standby Liquid Controt System Modifications Saunary l
This modification replaced the standard boron solution in the standby liquid control (SLC) system with an enriched boron solution to meet anticipated transient without scram (ATWS) requirements. As a result of enriching the solution. the heat trace piping was deleted, the storage tank heater controls were revised and the required tank level and a\\sem6 were revised.
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Sefaty Evaluetten Summary:
1 The functional capability of the standby liquid control (SLC) system has been increased without changing any of the key process parameters or the system's operation. Using an enriched solution does not change any of the key system process parameters (i.e., flow rates, discharge pressure, required not positive suction heat (NPSH). etc.) because the enriched solution is chemicaMy and physicaMy similar to the solution currently used. The changes are being achieved in accordance with General Electric's Topicot Report NEDE-31096-P-A.
This Topics 1 Report has been reviewed and accepted by the NRC. AdditioneM y, the SLC system is not taken credit for in any of the design bases accidents; it merely provides a backup to other safety-related systems. However, should the SLC system be esquired to perform its anticipated transient without scram (ATWS) function, the proposed changes win increase the functional mapability of the system such that the reactor wi n be shut down in approximately half the existing time. NRC issued a safety evaluation in support of Lioense Amendment No. SS which addressed this modification.
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DESIGN CHANGE SLSSMRY Safety Evaluation No.: 09-0053 UFSAR Section 7.2 Tables 7.2-2 & 7.2-3 n er. w.
laptementation Desument No.: EDP 10127 t
System No.: C7100 Title of Change Reactor Protection System Logic Modifications Sumunary:
This modification deleted the backup manual scram function of tripping power to reactor protection system (RPS) topics. Two additional manual trip togic channels identical to the two existing manual trip topic channets were added.
These changes were made to meet IEEE 279-1971 requirements for RPS manual contret single f ailure criteria and independence of the RPS manual / auto logic.
Safety Evaluation Summary:
The modification enhances the ability of the reactor protection system (RPS) to manuou y initiate a reactor scram by providing the redundant manual scram channels actuating a trip in each auto scram topic channel. Deletion of the backup manual scram (BUMS) system does not of f act the RPS topic trip functions.
Backup manual scram actuation was provided to de-energine the RPS power to both RPS trip topics A and S.
This in turn de-energised the nuclear steam supply shutdown system (NSS$$) topic pswer and caused an unnecessary closure of main steam isolation valves. Removat of the SLS48 trip topic wiu not af fecet the ability of the RPS to achieve saf a plant shutdown (scram), as origineMy designated. Therefers this meditt;ation wiu not atter the possibility of off-site or on-site radiologiset dose release. NRC issued a oefety evaluation in support of License Amendment No. 37 which addeoesed this modifiestion.
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EDP's Pope 21 DESIGN CHANGE StadbERY Safety Evaluation No.: 49-0089 Figure Change Implementation Dooumont No.: EDP 7838 Systen No.: A7100, 82103 Title of Changet Main Steam Isolation Valves Modifications Sesnary:
The manual control circuits of the existing inboard and outboard main steam isolation valves (MSIV) were modified. The modification replaced the momentary pushbutton actuated seel-in relays with latching type relays to better emulate maintained position control switch contacts. This was done to eliminate potential inadvertent valve operation due to recovery from loss of power to the MSIV manual control circuit causing drop out of the aest-in relays.
Safety Evaluation Sunsmary:
This modification anhanced the manual control of the main steam isolation valves (MSIVs) by providing a reliable means of manual control and at the same time eliminating the problem of inadvertent MSIV operation during the perf ormance of survein ance testing. The change in retey type and mounting location does not create a new failure. The f atture of MSIV to operate is stin the bounding case and is unchanged by this modification. The modification does not of f act the design bases or operational f unctions of the MSIVs in the nuclear boiler system as originauy designed.
Safety Evaluations EDP's Pape 22 l
DESIGN CHAM 3E SLbtWtY Safety Evaluation No.: 39-0062 Figure Change Septementation Doounent No.
EDP 10230 System No.: 72201 Titte of Change Fifth Moor Reactor Building Enclosure y
Summary:
A seven foot high mesh enclosure for the storage of tools was erected on the fif th floor of the reactor building west of the Wocontamination pit.
i Safety Evaluation Sesnary:
I The enctoevee is secured to structural concrete and structuret steel columns.
During a posulated seismic event no safety related equipment win be impacted because there is none close enough to be damaged during the event. The enclosure is located such that none of the existing fif th floor main crane activities nor laydown areas or heavy toad travet pathways wi n be impacted or l
hampered. The engineering design package imptomenting this modification I
accounted f or the change in structural loading. This modification did not doet with equipment modifications.
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Safety Evaluations EDP's Page 23 DESIGN CHANGE SLASWLY Safety Evaluation No.: 39 0063 UFSAR Sections 6.2 & 9.3 Imptementation Document No.: EDP 10105 System No.: P6000 T2200 Title of Change Reactee Building Railroad Airlock Door Modifications i
Smunary:
The purpose of this modification was to provide a qualified source of air f rom the non-interruptible control air system (NIAS) for the inflatable air esats on the reactor building restroad airlock doors, thus maintaining secondary containment integrity during design bases events. including site floods.
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Safety Evaluation Summary Secondary containment integrity is enhanced by this modification, ty providing a qualified source of air and a quotified means of inflating and maintaing the sea'.s assures the railroad air lock doors win perf orm their safety related function. Originau y technten specifications required only one railroad airlock door to be closed. The amendment to the technical specifications.
reference totter NRC-gg-00g3, requires both doors to be stosed under normat operation. With one division of non-interruptible control air system (NIAS) being supplied to the outer door and the othe,* division to the inner door, and with both doors closed, the system is single f atture proof.
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Safety Evolvations EDP's Pope 24 1
l DES 194 CHANGE SLAAMRY Safety Evaluation No.: gg-0073 Figure Change 1mplementation Deeumont No.: EDP 10166 i
System No.
E4100 Title of Change High Pressure Coolant Injection System Levet Switch l
Replacement l
Summary:
The tevet switch in the high pressure coolant injection (HPCI) turbine steam supply condensate drainpot was no longee operable and needed to be replaced.
The tevet switch was manuf actured by Robertshaw and supplied by GE: a Robertshaw direct replacement was no longer available from OE. The levet switch was replac$d with a unit manuf actured by Magnetrot under a OE purchase order. The l
recommended levet switch was dif f erent in f orm and fit (mechanical and otectrical connections) but its function in the HPCI system, as originauy designed, remains the same.
Safety Evaluation suunney:
Replacement of the levet switch maintains the OA levet, seismic category. ANSI and AshE classification of the previous owitch. The replacement Magnetrol levet switch was origina u y provided by GE for use on the OAL 1. seismic sategory 1.
piping group B scram discharge volume in the control rod drive system. The HPCI drainpot is also GAL 1. seismic category 1. piping group B.
The changes do e.ot ef fect the design besos or the intended function of levet switch in the k.*gh pressure sootent injection (HPCI) system nor de they alter the limiting senditten fee operation (LCO) requirements in the technical specification for the NPCI system er the description in the updated finat safety analysis report (UFSAR).
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DESIGN CHANGE SLAARRY Sefaty Evaluetien No.
89-0076 F1 pure Change 3mplementation Document No.: EDP 6740 System No.: 83100. C3200, C3600, H2100 Title of Change Reactor Pressure and Level Instrumentation Rock Repiscement Summary:
l This modification replaced two existing reactor pressure and level instrument racks with a completely redesigned version of each rock. This method of replacing an existing rock with a completely new rock was selected primarily to meet schedule requirements and Tech Spec operability restraints. The time period that the individual or conective set of instruments contained on the rocks can be out-of-service is limited to the period of time when not handling l
1rradiated f uel in the secondary containment, with no core alterations, or operations with a potential f or draining the vessel in operational conditions 4 and 6.
Safety Evaluatien Sununary:
Insta u stion of the reactor pressure and level racks, and instruments wi u not result in the increase of the off or on site radiologiont release since the racks and instruments are designed to meet or exceed an applicable Edison and NRC Regulatory requirements. The enhanced human f actors characteristics of the rock closign guarantees a decrease in the probability of a malfunction due to survel u ence testing. In addition, the single failure capability of the protective system is retained if not improved by this change.
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1 Safety Evaluations EDP's page 26 DESIGN CHANGE StaSERY Safety Evaluation No.: 39-0078 UFSAR Sections 5.2, 7.6, 7.7, 15.0 & 15.8 Figure Changes t
Implementatten Dooment No.
EDP g$42 System No.: 83100 i
Title of Change: Design Improvements of the Anticipated Transient without Scram Mitigating Systems Sumunary:
The alternate rod insertion (ARI) and recirculation pump trip (RPT) systems were modified to improve the design of the anticipated transient without scram (ATWS) mitigating systems. These improvements were in line with 10CFR50.62 requirements for AR! and RPT systems. The proposed modification fo n owed the recommendations of Licensing Topical Report No. NEDE-31096-P-A Anticipated Transient Without Scram Response to NRC ATWS Rule. 10CFR50.62". which has been accepted by the NRC. The actual modifications included a) manuel ATWS (AR1/RPT) initiation capability f rom the control room, b) esal-in functions of the initiating algnals with manual reset, c) testability at power.
Safety Evaluation Summary:
This enodification does not alter the safety function of the alternate rod insertion (AR1) and recirculation pump trip (RPT) systems which an ow safe reactor shutdown during anticipated transient without seram (ATWS) ovent. The modifications provide enhaneomont to the existing ARI and RPT systems which wi n add to the system's capabilities those fastures for the ATWS mitigating system
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which have been endorsed by the NRC (Ref e rence G. E. document NEDE-31086-P-A).
l Thereforo, this modification win not impact the radiologiset on-atte or l
eft-site deoes that would likely result free any accident. Adding the manuel j.
initiation espability from the sentrol esem for the AR1 and RPT systema does not affect the existing automatic initiation of the system. Also, eeneurrent manual initiation of AR1 and apt does not offact reactor shutdown espability etnoe operator's actions win be taken under emergency senditions when conteot eod l
insertion is warranted. Similarly, providing seal-in function to the ARI and RPT initiation logics (to assure completion of the ATWS function af ter the initiation signal is received) enhances the safety function of the system.
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DESIGN CDMNGE S4Aewty Safety Evaluation No.: 09-0061 UFSAR Table 7.8-11 j
Implementation Document No.: EDP 3011 i
System No.: 04100 Title of Change Fuel Pool Water Level Alarm Modifications Sammery This modification removed the ' Fuel Pool Water High Level" input f rom the ' Fuel Pool Cooling Trow' ole" alarm window in the control room and assigned the input its own alarm window and sequence recorder point. The alarm window
- Fuel Pool J
Water Low Level" was relocated to the lef t of the new window. The associated i
sequence recorder point for the low water level window will remain the same.
Safety Evaluation Summary:
1 This modification did not change the functional or operational configuration of
-.l the f uel pool cooling system. The changes are intended to more adequately
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apprise the control room operators of the system status, by independently atteming a fuel pool water high level, rather than sorporately alarming this eondition ae a pert of fwel pool cooling teouble. The changee defined by thie modification were for the alarm function only. There were no revisions to the control circuits of the fuel pool ooeting system.
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Safety Evaluation No.: 89-0005 Implementation Document No.: EDP E123 System No.: 82100. E1100, E2100. E4100. H1100 Titte of Change: Test Jack Revisions l
I Suunary:
Sanana jack type test points were instaued in certain relay room penets with each test point paraneling a corresponding point of the existing omergency core cooling system (ECCS) multipoint Jacks. Insteu ation of the now test Jacks was l
tow on the panels making them more readily accessible f or use than are the l
previous test jacks. The tower location wiu not require the use of a ladder J
l for test plug insertion, wiu not tend to stretch the test leads and win help prevent short circuited leads which have caused reportable events.
Safety Evaluation Suunary:
The design change instaued is safety related in that au new test points connect directly to existing safety related circuits. As each connection point paranels an existing point and is mounted in individuauy drined hotos, and is connected to the terminal booeds by individual wires, the only feasible single f ault f ailure is that of a single point shorting to ground. At worst, such a short could blow the fuse protecting a divisional power supply, thereby deactivating part of the protective circuitry. This condition stready exists in the present instaustion and is nueviated by the redundancy of the other divisional channet. Instaustion testing wiu proctude accidental grounds or ehert circuits.
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DESIDI CHANGE SLAMARY Saf ety EveWatten No.: 89-0087 leptementation themment No.
EDP $931 System No.1 E4100. E6100 Title of Change Limitorque Operator Spring Pack Replacements Sununary:
This modification entailed the replacement of Limitorque operator spring packs
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for valves in the high pressure coolant injection (HPCI) and reactor core injection coolant (RCIC) systems. This change enhanced the current valve actuator perf ormance by reducing the r*4 Wired spring p6ck compression, thus.
I altowing f or a higher working band to of f set thrust losses during operation.
Safety Eve hation Summary:
The modification provides an enhancement of current valve actuator perf ormance
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for motor-operated valves (MOVs) in the high pressure coolant injection (HPCI) and reactor core injet: tion coolant (RC1C) systems. The required stem torque is
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obtained at a lower torque switch setting. This reduces the susceptibility to spring pack hydraulic damping that could cause damaging overthrust. The heavier duty pack also increases the available maximum torque capability of the actuator to compensate for future valve wear.
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DESIGN CHANQE SLBARRY i
sef aty Evahaation No.
89 0114 Figure Changes l
!aptementation Docuument No.: EDP 10518 System No.: E1156 Title of Changes Fan Motor Brake Design Evaluation Sununary:
This EDP modified the Mechanics 1 Draf t Cooling Tower f an motor brake design to bring the design into regulatory comptience. The mechanical draf t cooling tower f an motor brakes are provided for tornado protection. The changes made included qualifying to Quality Assurance Levet I and Seismic ! interposing relays, inverters, and brake sotonoid. The power supply to the f an brakes was changed to Class 2E. DC battery, tafety Evetuation Susumary:
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The modified design completely qualifies att the components of the Mechanical Draft Cooting Tower fan brake system to 0A1. Seismic 2 coquirements, and the a
divisional Class It battery power sources provide the redundancy which proctudes an equipment matfunction.
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DESIIDi CHAMQE Stee WtY i
e Safety Evaluation No.
09-0124 Figure Change r
lap h oontatten Desument No t EDP 10698 f
System No.: D1100 Title of Changes Flow Switch Repistement I
Summery:
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This modification replaced a flow switch associated with a decent radiation monitor instau ed under a previous modification. The original flow switch FCI 6
lAODEL No. FR70-4 series thermet switch did not function property due to alpee and dirt deposits. The new switch is a Serksdals Model D2H-H1888.
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Safety Evaluation Summary:
The replacement of the flow switch with 6 dif ferent type is intended to improve i
the rettability and performance by stiminating components that are ef fected by algae build-up such as the previous switches used for this instenetton. The function of the pressure switch (to provide stare in the main control room) l comains the same as origina n y designed and does not affact the performance of i
l' any equipment se component important to safety. The configuration change necessary to insten the new switch does not change the intent of the original design.
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i DES 10N CHANGE SLSSERY Safety Evoluetter No.: 89-0125 Figure Change 1
Implementation Doo ment No.: EDP 9666 1
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- System No.t N2102 J
Title of Change Reactor Foodpump Turbine Speed Control Loop Modifications I
t Suunary:
i This modification removed the position f eedback signal and other associated control system changes to the reactor f oodpump turbine speed control loop.
Safety Evaluation Samary:
Reactor speed control loop modifications reduce the probability of loss-of-f oodwater events by providing an improvement in f eedwater control system stability. The modification reduces the probability that a control system malfunction will result in a high reactor water level transient. Reduction of
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the probability of a high reactor water level transients reduces the challenges to the level 6 trip system.
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Safety Evaluetten No.: 89-0128 Figure Change Implementation sw amont No.: EDP 9962 System No.: N2100 l
Title of Change Feedwater Startup Flow Transmitter Recalibration l
Stanaary:
This modification changed the Control Room indicator seating to 0-7860 ppm f or the feedwater startup tina flow indicator. The scaling had been 0-4000 gpn.
i During tow power operation, bef ore transf erring feedwater flow to the normal l-lineup, startup flow wiu exceed 4000 ppm. Actuat transfer occurs at approximately $126 gpm. This higher flow is used because it results in more stable feedwater operation. A value of 7860 ppm was chosen since it ccrresponds to the maximum flow the existing transmitter can measure.
Safety Evaluation Sumanary:
The change provides operators the means to monitor feedwater flow during a critical evolution. Having the flow above 4000 gan ons per minute (GPM) in the foodwater startup flow line for short periods of time (estimated at less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year) win not impact the line or the startup level control valve. Total erosion on these components for the life of the plant win remain essentiauy the same. This modification rescaled existing equipments no new enuipment is involved. This transmitter is used for indication only during unit startup. This function remains the same. Actue u y having higher flow during transfer of control reduces instability problems.
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Safety Evaluations EDP's page S4 DESIGN CHANGE tubMARY Safety Evaluation No.
30-0137 Figure change Zaptementation Doomment No.
EDP 10624 System No.: T4500 Titto of Change Elimination of a Postulated Flooding Scenario Summary:
The design change addressed in this modification was to eliminate a postulated f\\ooding scenario of a non-interruptible sie supp\\y equipment room (N1AS) by b\\ocking floor drains in the reactor bu1\\ ding closed cooling water (RBCCW) heat exchanger room in the auxiliary building.
Safety Evaluation Summary:
This modification replaced temporary rubber plugs being utilised to prevent the non-interruptible air supply (N!AS) equipment room from being flooded during a postulated accident scenario. No safety related equipment, structur6 or system was modified by this design change. This modification win prevent potential flooding of the non-interruptible contret air room. This wiu keep the availability of non-interruptible air supply system equipment the same se peeviously evaluated in the updated finot safety anatyeis report (UFSAR).
e Safety Evaluations EDP's page 35 DESIGN CH8ME SLAAMRY Safety Eve Wation No.: 89-013g Figure Changes
!aptementation Dooment No.: EDP 10446 System No.: 72200 Titte of Champo Construction of a Reinforced Masonary Walt Swana?y:
This modification was issued to provide design and construction details f or the plecement of a seismic 21/2 reinf orced masonry well on the 3rd floor reactor building to provide suf ficient shielding f or the anticipated radiation f rom the fuel pool cooling and cleanup (FPCCU) system suction line (Reference Design Calculation (DC)-5078 Rev. 0).
Safety Evaluation Summary:
Addition of a matt wilt not ef fect the f uel poet cooling and cleanup (FPCCU).
system, nor will it have any interf ace with any saf ety related equipment. This wolt is a non-toed bearing partition wait and does not support any safety rotated items. The structural integrity of this weit for Sotomic II/I criteria
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is being documented in Design Calculation DC-SA1 Vol. !!! DCD. Addition of the shioid watt f or the FPCCU system suction line is required ' pro-ALARA" to I
atteviate potentist future radiation problems. It will reduce dose rates to personnet in the als\\o way adjecent to tbs 11nse.
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DESIGN CHANGE SLASERY
-l Safety Evekstion No.: 09-0142 1sytementation Deoumont No.t EDP 10237 System No.: C9100 Title of Chance: Nuclear Steam Supply Software Changes Siammery A new nuclear steam eupply (NSS) NFD/PC software model on the plant process computer was instaued to support fuel design 6.
This software changed the methof for sletermining the power distribution and thermal limit calculations.
Safety Eva%ation Stannary:
The ef fect of the new NFD/PC sof tware model on the plant process computer is 11eited to tht nuctose steam supply (NSS) software and NSS dstabank. The only
. changas are software changes which have been shown to produce power.
J1stributisne and thermal margies comparable to that of the previous software.
- The new sof tware has been tavorably compared to the previous sof tware and was property instaued and tested with a detailed test procedure.
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DE8194 CHANGE SLAAMRY Safety Evakastion No.: 49-0150 laphunentation Doousent No.: EDP 10514 System No.: T4102 Title of Change: Access Doors Restricting Devices 46sumery:
The control center complex air quality is controued by various equipment (e.g.,
fans.. filters, heaters, humidifiers) located mainly on the fourth and fif th l
floors of.the aux 111ery building (AB), Control center heating, ventitating, and l-air conditioning (CCHVAC) ducts direct the air between the equipment and control room.. Numerous penets and doors in the ducts a Mow access to the duct' interiors for maintenance and testing of equipment and dampers. The access doors and e
panels were not provised with access restricting devices and could thus s n ow unauthorized personnet into the control room via the 6xisting access doors and penets and then through the ducts. The new design consists of permanent tocked barriers insta ued at 13 locations in or near the CCHVAC ducts - two inside duct penetrations and etoven outside duct access panels and doors.
Safety Evaluation Summary:
The addition of the barriers win not af fect any previously evaluated accident conditions. The barriers act to restrict access to the control conter complex i
and do not alter the safety function of any system or equipment. The expanded' metal barriers inside the ducts do not adversely restrict mir flow and do not obstruct fittore or damper operation. Thus the required flow paths for an
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modes of control conter heating, ventitating and air conditioning'(CCHVAC) operation, both normal and emergency, are maintained and the radiological eenesquences to the operators, as oveluated in the updated fir 41 aafety analysis esport (UFSAR), are not affacted.
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Safety Evaluations EDP's Page 38 i
DES 10N CHANGE S&SAMRY t
Safety Evaluation No.: 89-0166 Figore changes s
Implementation Docueent No.: EDP 9724 3ystem Mo.: N2200 3.
Title of Changer Emergency Drain System Valve Woc!ification t
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This modification instaued a 3-way air operated pilot valve on the operating air line of. level controt v.tves which are part of alternate emergency drain systems associated with feedwater heater 5 north and 5 south respectively, so that the operatira air line run to the valves win be shortened and the control response win be improvec, Refety Eval.uation S.sunary:
The propcsed modification ef fects only the operating air source for the l~
operation of the emergency drain control, valves. which are part of the heater
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drains system. The basic operation and control initiation of the control valves are not affected,. To enhance the control, response of the drain control-valves to the-levet eighets, new 3-way air operated pilot valves are added in the operating air supply lines. Addition of these new components in the feedwater heater drain system wi n improve the performance of the affected foodwater
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beater drain valves. Failure of the feedwater heater drain valve win not affect any safety related equipment or system.
Safety Evalu6tions EDP's Page 39 DESIGN CHANGE SLANARY Safety Evahastion No.
89-0157 b? TAR Table 3.10-3 laplasmentation Document No.: EDP-1D C System No.:.E11. E21. E41 Title of Change Replacement of Relays Susunary:
EDP 10467 replaced eleven existing HGA relays which use the normany closed contact in the de-energized st&te f or saf ety related f unctions, with seismicany gustified HnAA relays. Tbs >tAA relays are seismicauy rated f or 2.5g ZPA. The reteya are used in control circuits for RHR pumps and in various RHR, core spray and HPCI system val.ves. This change resulted from NRC Inf ormation Notico 88-14.
Safety Evaluation Susunary:
The M relay is a proven, rollable component, and is an acceptable qualified
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replacement for the HGA relay. The f unctions and the logic of the pump motors and the valve operators, in the relay circuits, wi u not be affected by this modification. The margin of saf ety in LOCA analysis, which is the basis for ECCS Technical Specifications win not be ef f ected, because the >SAA relay is a functional equivalent of the HGA relay.
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. page 40 DESIGN CHANGE SLASAARY Safety Evaluation No.: 89-0165 Zeptementation Document No.: EDp-10225 System No.: 32106 Title of Change Instaustion of Snubber Fitting on Differential Pressure Transmitter Sunnary:
EDP 10228 instaued a qualified instrument damping devin (snubber fitting) on-the high side of presspre dif ferential transmitter. 821N487, to eliminate pressure surges being sensed by the transmitter. 921N487 is used to sense difforential pressure between the upstream side of the third MSIV and inboard MSIV. The transmitters output is used by the Division.I MSIV leakage controt system to control the position of valve F436. This valve provides air pressure to help-prevent leakage from the MSIVs. Because of pressure fluctuations in the steem line the valve was beAng cycl.ed unnecessarity. Addition of the snubber on the dp transmitter eliminated this, problem.
Safety Evaluation Summary:
The ISIVLCS saf ety f unction is not af fected due to the addition of the enubber fitting. Adding the snubber fitting improves the performance of the transmitter by eliminating flow peessure auegos and eliminates cycling of valve F436. The enubber fitting is a passive and rettable device which is qualified for esfety
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related applications. The snubber fitting does not af foot the operability and availability of the system. Theref oee, the margin of safety as defined in the Technical, Specifloation bases 3/4.8.1.4 is not reduced.
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-Safety Evaluations EDP's Page 41
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DESIGN CHANGE SLAS4ARY i
Safety Evaluation No.t 89-0166 Implementation Dooumont No.: EDP 10757 System No.: 82100 Title of Change Level Transmitter Change Sumnary:
This modification replaced the existing CA Level 1. Seismic Category 1 Rosemount e
Model 1153 D85RC levet transmitter in the nuclear boiler (NB) system. (reactor pressure vessel water level - wide range) with a OA Level 1. Seismic Category 1 Rosemount Modet 1153D85RCN0037 transmitter.
' Safety Evaluation Susanary:
This modificatica replaced the f ailed Rosemount transmitter with a qualified l
Rosemount transmitter manuf actured under an isapecved process. The replacement l
transmitter has previously been environmenta u y and'oeismica u y qualified for'
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L-use in monitoring reactor pressure vesset water level. There are no as-tow-as l
reasonably achievable (ALARA) concerne for this modification and no permanent shielding is required. There are no human factors concerns based on this transmitter replacement. The plant fire protection program is not affected.
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. DESIGN CHANGE SLASMRY Safety Evaluation No,t 89-0167 Figure Change H
Implementation Document No.: EDP 10778 System No.t-N2200-Title of Change Valve Modification Sunnary:
The purpose of this modification was to insten a blank plate inside the body of a south separator seat tank. This valve had been damaged beyond repair.
Specificany. the seat ring and disc were damaged. This valve was instaued in the moisture separator seal tank drain line to the condenser to bypass and 7
protect a level control valve during startup and low power operation.
Safety Evaluation Suunnery:
l The moisture separator seat tank drain line. function and operability are not l
affacted by the change and furthermore the instau ntion of the blank plate does not affect.the accident er transient analysis..The instaued plate and specified weld were sized based upon system design conditions. The proposed change does not affact any equipment integral to plant normat operation or safe shutdown.' Since the negative effects of the damaged valve neekage) are removed overen system reliability and performance are enhanced. The separator-seat tank drain function is maintained.
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EDP's Page 43 DESIGN CHANGE SLANARY Sefety Evaluation No.: 89-0176 Implementation Document No.: EDP-10777 Syetom No.
T5000
-Title of Change Rerouting of Drywen Atmosphere Sample Line Summary:
Sample point CT T50 L4008 is one of the five sample lines in Division 2 that provide drywe n atmospheric samples to e. hydrogen and oxygen sensor. Sample point CT T50 L400B draws its sample from above the reactor pressure-vessel, head. The sample line associated with CT T50 L4008 above the buW head seal plate is supported from the reactor pressure vesset head insulation support structure. While lif ting the drywen head during the 1st refuel,ing outage. the innee ring of the head caught on a support, damaging the support and the sample
'line. 'The close' tolerance between the inner ring of the drywen head and the support makes the possibility of future similar occurrences likely. The change
- made by EDP 10777 is to reroute tho sample line to minimize the possibility of future damage.
Safety Evaluation Stannary:
The modification does not af f ect the operation of the system or 'its ability to monitor the Containment Atmosphere, since the change in pipe length is insignificant and does-not change sample flow rate or sample point location.
The modified sample line routing is designed and instan ed to Seismic Category 1 criteria. Changing piping material from Type 316L to Type 316 or 304 wi u not
- alter the corrosion properties of the sample line.
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Page 44:
DESIGN CHANGE SLAMARY Safety Evaluation No.: 89-0179 Figure Change
' Implementation Dooumont No.: EDP-10448 Systen No. - 03300 Title of Change: Removal of RWCU System Pipe Flanges u
Summary:
Flanges in the'RWCU piping between the return flow element FE-G33-N040 and the l.
return flow check valve 033F120 were removed and replaced with'4" pipe. These ly flanges were originauy supplied for hydro but otherwise provide no function to-
- RWCU, l
Safety Evahastion Summary:
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l The removal of the flanges eliminates a potentiat Teak-pathr thus chal,lengea to --- ---~~----
temperature and differentist flow instrumentation, and isolation valves wil.1 be further minimized. Furthermore, the design bases, functions, or operation of the RWCU system is not impacted.
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Safety Evaluations EDP's Page 45 DESIGN CHANGE SLANARY Safety Eval.uation No.: 89-0186 Figure Change Implementation Document No.t EDP 10840 Systom No.*
R1400 R1600 Title of Changet Swing Bus Motor Control Center Modifications Summary:
The 480V swing bus motor control conter (MCC) 72CF provides power to critical residual heat removal (RHR) and recirculation system leads. Thwst loads are the Recirculation pump discharge valves, and RHR injection and crosstie valves.
This modification enhanced the existing system. designed to detect a degraded voltage condition (minimum voltage esquired to ensure operability of swing bus loads) after EDG breaker closure under the accident scenario of loss-of-coolant-accident (LOCA) simultaneous with loss of of f site powee (LOSP) and a subsequent EDG (12 or 14) excitation system f ailure during the RHR pump motor starting sequence (e.g., vol,tage does not recover from starting transient). This degraded voltage detection win initista an automatic transfer of the swing bus to the unaf f ected divisional power source.
Safaty Evaluation Summary:
This modification is f or a scenario in which the voltage regulator f ails to a degraded voltage condition after closure of emergency diesel generator (EDG) #12 or #14 output breaker (e.g., residual heat removal (RHR) pump start wiu dip voltage and never recover). Further in this scenario, the normal, owing bus throw over wiu not occur since the regulator did not completely fait.
Therefore, an undeevottage (W) device with five (5) second time delay (TD) relay was added to detect the degraded vot,tage. Atter a postulated design basis accident (DBA) and lose-of-of f eite-power (LOSP) with the single event f ailure of EDO 912 er #14. there win be 2 core sprey pumps. 31.sw pressure coolant injection (LPCI) pumpe, high peassure soolant injection (FPCI) and automatic depressurisation system (ADS). The swing bus at Fermi 2 is provided for LPCI function of RHR. The most limiting single f ailure for LPC1 initiation is the DBA and LOSP with' fail,ure of Division !! battery. (The postul,ated W failure on EDO 12 or 14 under the existing design is not eensidered a eredible event and is not part of the original design basis). This osse toeves only 2 core spray pumps 2 LPCI pumps, and ADS system. This event is more conservative than the proposed modification scenario since it han one less LPCI pump and no HPCI system.
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-Safety Evaluations EDP's Page 46 l
l DESIGN CHANGE SLASERY Safety Evaluation No.: 89-0196 UFSAR Table 5.2-6 Figure Change l'
Implementation Deounent No.: EDP 10792 System No.: B2100 l
l-Title of Change Reactor Head Vent Valve Modifications
%sumery:
Specia u y designed fittings are instaued immediately upstream of the. reactor vessel head target rock, pilot actuated solenoid valves (SOVs). These valves were'insta ued to ensure the reactor coolant pressure boundary (RCPB) since the-target rock valves were prone to developing leakage through the pilot / seat.
Refety Evaluation Summary:
This change does not have any impact on the plant fire protection / detection programs, nor on the physical security plan. While some negative impact on es-low-as-reasonably-achievable (ALARA) may. result from the need to vent the 1"
reactor head manusu y, this is an infrequent operation, normany required only l_
prior to reactor head removal or reactor hydrostatic testing. Eliminating the
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adverse local environmental impacts (i.e.. increased temperature and radiation).
the increased influent into -the radweste system, and the need to f requently repair the target rock valves is judged to more than of f set the negative ALARA l
impact associated with inf requent operation of the manual head vent valves. - The
'spectany designed fitting" was intentionauy derigned to miele the material and goemetry of the piece of pipe it replaced as much as possible, so that the response to system transients would be essentiau y the same. This "speciau y i
designed fitting" is qualified in design calculations. Thus reactor coolant peessure boundary (RCPB) intagrity ie enhanced with a minimal impact on systom configuration or en the ability of the system to meet its design bases.
E W OF EDP SECTION
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I FERMI 2
- SAFETY EVALUATION
SUMMARY
REPORT t :-
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1989 i
POTENTIAL DESIGN CHANGES '-
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i Safety Evaluations PDC's page 2 POTENTIAL DESIGN CHANGES (PDC'S)
Safety Evaluation No.:
87-0306 Y
Implementation Document No.: PDC 7209 System No.: N3017 Title of Change: Turbine Bypass Valve Fast Open Circuit Modification Sununary:
A jumper was added which shorts the dropping resistor in each of the two
" fast open" solenoid circuits of the turbine bypass valves. The effect of shorting the resistors is to allow the fast open solenoid to operate more quickly, and thereby reduces the time for the turbine bypass valves to open to less than 300ms (a Tech Spec limit).
lI Safety Evaluation Summary:
l' The original vendor design catted for the resistor to be in the circuit to
. extend the life of the solenoid. However, the fast open solenoid is only energized for one second in response to a. reactor scram. Thus, the effect of removing the resistor on solenoid life is-negligible. -The change was l.
needed to allow the turbine bypass valves to operate within Tech Spec l
required limits.
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Safety Evaluations-PDC's' Page 3 Safety Evaluation No.: 88-0029 UFSAR Table 3.8-39 Implementation Document No.
PDC 7894 System No.: E1100 Title of Change: Revision of Code Computed Stress Tabulation Sunnary:
This minor modification added permanent lead shielding blankets to the residual heat _ removal (RHR) containment spray piping near the c'ontainment penetration. An analysis of this modification showed that the modified piping system satisfies all applicable design requirements. However, the piping stresses tabulated in the UFSAR-have changed as a result of this design modification.
Safety Evaluation Summary:
Analyses have shown that the mounting of the lead blankets to the piping is adequate and that the piping system is acceptable considering the additional weight of the lead blankets. The ASME code computed stresses are tabulated in the UFSAR so as to provide sample representative piping stress data. Paping stresses in the modified system have changed f rom the values reported in the UFSAR but they still satisfy ASME code allowables.
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Page 4-
.l Safety Evaluation No.: 88-0061 Implementation Document No.: PDC-8398 System No.: R1400, R1600 R3100 Title of Change: Changes to Fuse Size and Breaker. Trip Setting-Summary:
This change revised.the fuse size from 15 to 5 amp for two 120Vac distribution circuits and reduced breaker long time trip settings for five 480Vac breakers. The purpose of these changes was to reduce the available
' fault currents so that fire induced high impedance faults will be unable to
-challenge the power sources needed to provide a means to shut down the plant in the event of a fire. The changes resulted from a study done to determine compliance with 10CFR50, Appendix R as additional guidance was provided in NRC Generic Letter No. 86-10.
Safety Evaluation Summary:
The changes as described to the brer_k :
eip points and fuse ~ sizes provides sufficient margin between the connec!<<.oads and the points of operation of the protection devices as to assure operability under normal (non-overload /non-short circuit) conditions, retains coordination, and provides overload /short circuit paotection in the event of a fault or equipment malfunction, t
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~ PDC's Page 5 Safety Evaluation No.: 88-0085 UFSAR Table 7.6-2 Implementation Document No.:
PDC 7578 System Ho.
T5000 Title of Changes Primary Containment Hydrogen Monitor Setpoint Change Sunnery:
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The lower alarm setpoint f or the primary containment hydrogen monitor was reduced f rom 2 percent to 1 percent hydrogen concentration to provide more margin to the 6 percent analytical limit and to match the emergency operating procedure (EOP) action requirements to be taken at 1 percent hydrogen concentration.
Safety Evaluation Summary:
Design calculations show that the new alarm setpoint of 1 percent "Div I/Div II Hydrogen Concentration High" is more conservative when compared I
with the high-high atorm setpoint of 3.5 percent and the analytical limit I
of 6 percent. Selection of a more conservative setpoint improves the primary containment hydrogen monitoring system by providing more margin to safety. Selection of a more conservative setpoint does not affect the basic function or design basis of the system.
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Page 6 Safety Evaluation No.:
88-0154-Figure Change
' Implementation Document No.: PDC-9304 System No.: E4100. E5100 Title of Changer HPCI and RCIC Drain Pot Piping WWterial Changes Summary:
This minor modification was written to provide design details, material and l
instructions necessary for the replacement of damaged carbon steet piping l
and components (isolation valves) located near the condenser in the high f
pressure coolant injection (HPCI) and reactor core isolation cooling.(RCIC)
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system drain pot drain lines to a condenser nozzle with stainless steel materials.
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Safety Evaluation Summary:
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The change essentially was a'materlat change only implemented to increase system resistance'to erosion and pitting damage. The isolation valves were j
replaced with valves that are identical except for material (stainless) l]
and, except f or 45 elbows being substituted for 90 elbows (as flow l
enhancement) the piping configuration remains identical. The materlat replacement-is suitable for the design conditions which will be encountered j
-and the impact to the system stresses reviewed and determined to have
, negligible effect. This change was considered a plant enhancement and was
_j j-implemented to increase plant availability /rettability. The portions of l
L the high pressure coolant injection (HPCI) and reactor core isolation
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cooling (RCIC) systems that were replaced are not safety related and are
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. not identified in the UFSAR safety analysis.
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Safety Evaluation No.: 88-0160 Figure Change Implementation Document No.: PDC 9359 System No.:- N2200 Title of Change Removal of Drain Line from MSRs Summary:
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This minor modification removed piping, associated valves and capped the lines to a. drain line from the Moisture Separator Reheaters. The drain line was originally instatted to be used to drain condensate from the MSRs i
when building heating steam was used to blanket the MSR while shutdown, L-However, because of concerns for cross contamination, building heating steam will never be used, and as a result the drain piping was not needed.
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Safety Evaluation Susanary:
J This change has no effect on the operation or function of the MSR.
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Safety Evaluations PDC's-Page 8 Safety Evaluation No.: 88-0167 Figure Change Implementation Document No.: PDC-9289 System No.: R3200
- Title of Change ' Balance of Plant Battery Charger Setpoint Changes Summary A tolerance of + 0.5v was added to the starm and trip setpoints for the division I, 11 and balance of plant (BOP) battery chargers, the high voltage alarm setpoint for'the BOP chargers was changed and the float voltage settings for both divisions and the BOP chargers was decreased from 135.0 + 0.5v to 133.0 1 0.5v.
In 1984 acid was added to the batteries which increased the specific gravity to greater than 1.250 and required an increase of the float voltage; prior to the change the float voltage was 132.0v.
In 1986 new batteries were insta11ed necessitating the changes described above.
Safety Evaluation Summary:
The addition of a + 0.5v tolerance to the alarm and trip setpoint limits for the division I and II and BOP battery chargers, changing the high voltage alarm setpoint f or the balance of plant (BOP) chargers, and the
-reduction of the float voltage from 135.0 + 0.5v to 133.0 + 0.5v for the division I,-II and BOP chargers are within the specified limits of the UFSAR. plant technical specifications, and the vendor operation instructions documented in design calculation (DC) No. 0539, revision D.
Safety Evaluations PDC's.
Page 9 Safety. Evaluation No.
88-0178 Implementation Document No.: PDC-9428 System No.
C3200 Title of Change: Deletion of Reactor Feed Pump Runback on Loss of Heater Drain Pump Flow Summary:
The f eature that provides a reactor f eedwater pump runback in speed when a toss of heater drain pump flow is sensed was removed. The heater drains pumps normally provide 1/3 of the suction flow for the FW pumps and the remaining 2/3'is provided by the heater feed pumps. When heater drains flow is lost, recirculation pump speed is lowered resulting in a reactor power reduction. The reduced reactor power level results in a reduction in FW demand, thus ensuring the FW pumps will not trip on low suction pressure.
Safety Evaluation Summary:
This change will provide more margin in avoiding reactor water level.3 scrams during a loss of one FW pump transient, thus avoiding unnecessary challenges to protection systems. The FW pumps are still protected from a trip.on low suction pressure by the recirculation pump speed limiter when a loss of heater _ drains flow occurs. The physical change made to implement deleting the FW pump runback on loss of heater drains flow was accomplished by-resetting existing FW control equipment, thus no new equipment faltures were introduced. The failure consequences of the device are not changed and are bounded by the toss of FW transient evaluated in the UFSAR. This
' feature is not addressed.in the bases for any technical specification.
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l Safety Evaluations-PDC's
- Page 10-Safety Evaluation No.: 88-0202 Figure Change
' Implementation Document No.: PDC-9353 System No.: H1100-
.i Title of Changer Substation Name Change.
Summary:
The name of the Brownstown No.~1 120KV line was revised to Swan Creek line on the control operating panel (COP) and one line diagram. On the COP, the nameplate engravings on the telemetering panel were changed.
Safety Evaluation Summary:
The nameplate engravings on the control operating penet (COP) are classified as OA III, non-seismic since they are very small and lightweight. The nameplates will not. alter the seismic qualification of l.
the COP.
The revised engravings are specified to the character standards
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for the Fermi 2 COPS to maintain the controt room consistency and not I
create a human factors concern.
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.Page 11_
safety Evaluation No.: 89-0020 UFSAR Section 9.2 l ';
Implementation Document No.: PDC-9992 System No.:.P4100
. Title of Change: General Service Water Low Levet Setpoint Change Summary:
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The general service water (CSW) system takes suction from Lake Erie. The take water is drawn into an intake canal, passed 'through a trash rack and traveling screen before entering the CSW pump pit. Low levet switches are installed near the intake to alarm on low take levels. This design document changed the. tow level setpoint from 562 feet (24 inches water column decreasing) to 569 feet (108 inches water column decreasing).
Safety Evaluation Summary Changing the low level setpoint from 562 feet to the higher 569 feet-is a conservative measure. By making this change the low level _ alarm will be received earlier and will permit operations personnet more time to take appropriate action.
END OF PDC SECTION
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l FERMI 2 SAFETY EVALUATION
SUMMARY
REPORT 1989
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L PROCEDURES, TESTS AND EXPERIMENTS
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Safety Evaluations Procedures, Tests & Experiments l'
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PROCEDURES, TESTS & EXPERIMENTS 1
Safety Evaluation No.: 88-0090 l-Implementation Document No.
HPP-23.800.04 1*
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Title of Changet Alternate Decay Heat Removat Methods i
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Summery l
l This procedure was written to provide alternate methods for decay heat j
removal when in cold shutdown. The procedure provides for use of the RWCU system in the blowdown mode to the condenser or radweste systems with makeup water supplied by either the standby feedwater via the condensate sterage tank or condensate via the condenser hot well. The procedure also attows operation of these systems witn certain automatic isolation features disabled which is allowed by Technical Specifications.in Mode 4.For RWCU att isolations are defeated with the exception of Reactor Water Level 2 j
and for SBFW the Reactor Water Level 8 isolation is defeated.
Safety Evaluation Summary l'
UFSAR Section 15.9 analyzes a loss of normal RHR shutdown cooling. The i
consequences of sur.h an event is the necessity to use other shutdown l
cooling methods to re-establish decay heat removal capability. Procedure NPP-23.800.04 provides two alternate methods of decay heat removal, and j,
i requires that at least one Reactor Recirculation Pump or at least one loop I
of RHR shutdown cooling is operating. Both alternate methods have i
sufficient heat removal capacity to adequataly remove all decay heat. The procedure defeats all RWCU system isolations except RPV Water Level 2.
However, the procedure requires that the Reactor be in Mode 4, and the RWCU l
system isolations are only required for Modes 1, 2 and 3.
The procedure also defeats isolation of the SDFW System on RPV Levet 8.
However, in Mode
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4 RPV Levet is operator controtted and may exceed RPV Level 8.
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Page 3 Safety Evaluation No.: 88-0238 Implementation Document Nn.: SOE B3103-88-01 System No.: 83103 Title of Change Reactor Recirculation Motor Generator Shutdown Summary:
Maintenance was performed on one reactor recirculation (RR) motor generator l
(MG)-set (which requires that the individual MG set be shut down), while the other RR MG set operated at reduced speed, thereby placing Fermi 2 in a temporary single' loop RR power operation. Technicat Specifications require that with one RR loop inoperable, actions are immediately initiated to-j reduce thermal power to less-than or equal to the limit specified in the
' Tech Specs within two hours and measures are initiated to place the unit in at least hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Sequence of events SOE B3103-88-01 proceduralized single RR toop operation within the twelve hour time frame.
Safety Evaluation Summary:
Performance of sequence of events (SOE) 83103-88-01 in single reactor recirculation RR loop operation does not remove or alter any systems required to mitigate the effects of accident / transients evaluated in the
--UFSAR.
Although the SOE disconnects a logic function (RR. runback #3), this is accomplished deliberately to decrease the probability of thermal instability defined by SIL-380. Performance of SCE B3103-88-01 in single RR toop operation operates outside of the region of instability defined by
.SIL 380 with 80% rod line operation. Thus the concern for thermal-hydraulic instability is minimized. Single RR loop operation has been demonstrated, successfully, in the loss of one RR pump test-STUT.06B.030. " Lessons Learned" from the STUT.06B.030 have been applied to the development of SOE B3103-88-01.
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' Safety Evaluations Procedures, Tests & Experiments Page 4 Safety Evaluation No.: 89-0031 Implementation Document No.: 24.405.03 System No.: T9200~
Title of Change: Secondary Containment Integrity Test Modifications Summary:
Surveillance procedure 24.405.03, seconcary containment integrity test. was modified to include testing secondary containment with the railroad car door seal deflated and permitting the test to be performed while the plant is in operational modes 1 and 2.
Safety Evaluation Sumary:
This test-was completed by starting the standby gas treatment system (SBGTS) while the-reactor building heatina, ventilation and air conditioning (HVAC) was in operation and then shutting down the reactor r:
building HVAC system. This ensured that secondary containm6nt would be maintained at a negative oressure and minimized the possibility of an i
unmonitored release. Failure of this test would have caused the plant to i
enter into a four.(4) hour limiting condition of operation (LCO) that i
i required the plant to be in hot shutdown within twelve (12) hours and cold shutdown within twenty four (24) hours. No new equipment or failure
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mechanisms were introduced to complete this test.
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Safety Evaluation No.
89-0039 Japtementation Document No.: SOE 89-12. ECR-10105-4 System No.
p5000, T2200 Title of Change: Raitroad Airtock Doors Water Leak-Test Summary:
Modifications had been completed to the reactor building airlock doors to ensure they were " watertight." To confirm that the "watert10ht" requirement had been met, this water leak test was performed.
Safety Evaluation Summary:
This test would not result in any water teakage that might affect plant components. No safety related components are located in the airteck area.
This test mainly af fected the area around the airtock doors. Leakage through the inner door seat was expected to be smatt. In the event of the failure of the reactor building side inner door dike, the Itoor drain on the first floor would carry the watea to the corner room sumps. The smelt amount of water would not flood the corner room (s).
In addition, the corner room sump pumps wers operable during this test.
t Safety Evaluations Procedures.-Tests & Experiments Page 6 Safety Evaluation No.: 69-0046
' Implementation Document No.
23.710 System No.: F1600 Title of Change: Refuel Bridge Interlock Modifications Summary:
Procedure change to 23.710 " Fuel Handling System," was implemented to defeat refuel bridge interlocks to permit testing of the refuel platform.
Safety Evaluation Sumary:
Tho' purpose of the refuel bridge interlocks are to prevent inadvertent criticality wheti the reactor is in the refueling mode. The procedure change which defeated the interlocks specified that the fifth floor reactor cavity secondary shield plugs must be in place and the plant must be in modes-1, 2, 3 or 4 before, and while, these interlocks are defeated. While in compliance with these conditions, the interlocks are redundant and can be def eated since the plant. is not in the ref ueling inoos and there can be no physical interaction between the refuel bridge and the reactor vessel.
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Eafety Evaluation No.: 89-0069 j
t implementation Document No.: ARP 4010 t
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System No.: N3000 Title of Change: Turbine Trips Defested i
Summery:
Alarm response procedure was modified to reflect that the main turbine was in operation with the vibration trips continuously def eated.
Safety Evaluation Summary:
The probability of an accident previously described in the UFSAR is not increased. The bounding accident would be the turbine missile accident described in the UFSAR and the consequences would be identical. The probability is not increased since currently a trip would be manually perf ormed, rather than automatically, wolt bef ore danger of a missile due to high vibration. The current administrative trip limit is defined in alarm response procedure (ARP) " Main Turbine Vib High-High."
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i Safety Evaluation No.: 89-0070
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t Imptomentation Document No.: NPP-24.402.07, Rev. 20 System No.: 72300 Title of Change Drywell to Torus Bypass Leakage Testing Sumary:
A new procedure was written to perform drywell to torus bypass leakage i
testing during operations. This test was required per Technical Specification Surve111ances to a\\ tow continued operation and provide assurance of proper function of the drywelt to torus vacuum breakers during the time that one of the Vacuum Breaker Closed Limit Switches was not e
operab\\e. The dif f erentist pressure used was 2811* water column (WC).
Safety Evaluation Summary:
The test pressure allowed by this procedure is bounded by an analynis that calculates containment peak pressure and loads. The drywell to torus vacuum breaker valves tested by this procedure are capable of withstandin0
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differential pressure conditions seen under the design base accident.
These conditions bound the required differential pressures of this test procedure. Primary containment sampling system is also disabled during this test to prevent falso indication of leakage. The hydrogen and oxygen monitoring functions can be disabled f or up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> por Technical e
Specifications. The primary containment atmosphere gaseous radioactivity monitor will alta be disabled to ensure the pressurization occurs without interference. This function will be accomplished by daily grab samples in accordance with Technical Specifications. The procedure test conditions do not depart from drywell or torus pressures allowed by Technical Specification 3.6.1.6 therefore no reduction in safety margin occurs for drywell or torus pressure.
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-i Safety Evaluations Procedures. Tests & Experiments Page 9 Safety Fveluation No.: 89 0072 Imr.tamentation Document No.: DCR 69-0430 System No.: G1100 Titto of Change Water Transf er Sunsmary:
Approximately 10.000 gallons of water procwased at Fermi 1 were transported i
to Fermi 2 for further processing and eventvat e-euse.
Saf ety Evaluation Sunenery:
The transport of the Fermi 1 (CF1) processed wuter to Fermi 2 (EF2). the processing of this liquid. 6nd evsntual eeuse of the liquid at EF2 will not increast ths probability ce consequences of an accident previously evaluated in the UFSt.R.
The accident associated with the discharge of radioactive ef fluent 1 described in the UFSAR involves the f ailure of both the andweste tanks &nd building. This analysis was designed to show that even under this very conservative scenario EF2 would not diset'sege redanactive ef fluents in excess of those specified in 10CFR20. Appendix Fu Table !! Colume. 2.
The tronsport, receipt 'and procest.ing of the EF1 11guld at EF2 does not negatively impact this analysis as the isotepic concentrattent are already below those specified in 10CFR20. Appendix B.
Table 11. Column 2.
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1 Safety Evaluation No.: 89-0080 l
Implementation Document No.: SOE 89-00
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System No.: T2200 Title of Change Reactor Building Water Intenkage Test Sunenary ?
Sequence of Events (SOE) 89-08 was written to perform a test to determine the water leakage during a postulated site flood condition through the outer 6oor of the reactor building railcar airlock with the inflatable seele in the deflated position and with passive seat material installed as weather-stripping per a previous plant modification.
Refety Evaluation Summary:
No plant equipment was affected by this test. During the test, ficod l
protection was assured by the placement of sand bags against the outer l
airlock door. Per SOE 89-08, the test could only be run if lake water l_
level was below 580.0 feet. Therefore, no site flood conditions were required to be evaluated for the duration of the test. No safety related l
components are located in the airlock area. The ef fect af the test on the l
ability to provide air tightness for secondary containment integrity has I
been demonstrated by the Engineering Research Department lab test on the Armaflex material and actual flood protection was assured through the use of sand bags against the outer door.
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Safety Evaluation No.: 09-0082 UFSAR Section 9.2 I
Implementation Document No.: NPP-23.420. Rev. 9 t
System No.: E1100 i
Title of Changer Residual Heat Removat Complex Heating and Ventilation Damper Positioning Summary:
Procedure NPP-23.420 *RHR Complex Heating and Ventilation," was revised to prescribe a method of manually positioning dampers to maintain area temperatures required by the UFSAR using the engineering analysis provided in a minor modification.
Safety Evaluettun Summary:
The heating, ventilation and air condition 1r.g (HVAC) system will be aligned so as to maintain the room temperatures within the design basis requirements and there wilt be no ef f ect on any accident scenarios. A previous minor modification documented a seismic analysis of the c-clamps showing that they are acceptable and will not increase the f ailure probability of the dampers. Any damper failurf, would be bounded by the analysis in the UFSAR which indicates that should temperatures rise to alarm levels operators would take the necessary actions in accordance with I
alarm response procedures. The opere!.11ty matrices provide damper.
l positions that will maintain room temperatures with the design basis. No unevaluated type of accident is created, r
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t Safety Evaluation No.: 89-0083 Implementstion Document No.: DER 89-0219 System No.: T2200 l
Title of Change Reactor Pu11 ding Railroad Airlo o. Door Wbdifications f
Summary:
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This saf ety evaluation was written to justif y the interim Fermi 2 plant operation with the temporary conditions existent at the reactor building railroad airlock door, as retr.ted to flood protection requirements, until the implementation of Engineering Design package (EDp) 10105. This EDP used the plant divisional Non-Interruptible Air Supply (NIAS) System to provide CA (evel 1 redundant source of air for the inflatable seals on both i -
airlock doors. As reported in Deviation Event Report (DER) 89-0219, the current independent air supply system to the inf1 stable seals is non-0.
As an interim measure, sand bags were temporarily staged outside the airlock for use in case of a site flood.
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Safety Evaluation Summary:
The temporary use of sand bags for flood protection is acceptable under unusual circumstances as discussed in Regulatory Guide 1.102, Revision 1.
Additionally, published flood prediction data indicate that worst case water elevation for 100 year recurrence flood will not reach site grade elevation of 683.0 ft.
With the usa of sand bags and the extremely j
i improbably site flood scenario, equipment important to safety in the l
Reactor Building is adequately protected against flooding.
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Safety Evaluation No.
89-0091 Implementation Document No.
System No.: T4500 Title of Change: Inoperable Reactor Building Sump Cross Tie Flood Control Valves Summary:
This is an evaluation of the plant conditions regarding inoperable reactor i
building sump cross tie flood control valves. Deviation Event Report (DER) 89-0523 required att inoperable cross tie volves to be closed. This was done to preclude the cross flooding of corner rooms, and the high pressure coolant injection (HPCI) pump room, by virtue of drains backing up.
However, due to water build-up in the HPCI room from the equipment drain flows Operations used a portable pump to transfer water to the floor drain system. Due to the high frequency of draining this area and the high radiation levate in the sump. Operations requested a review of the acceptability of leeving cross tie valve between the HPCI room and the Division II RHR corner room open. The result of the review was that the valve could be left open.
Safety Evaluation Summary:
The function of the anti-flood cross tie valves is to close and prevent cross flooding. Since all valves except the high pressure coolant l
injection (HPCI) cross tie valve are operable or closed. and since this valve is closed if pending flood conditions exist. no now flood scenarios are created. Further, the analysis of the consequences of a safety equipment malfunction are unchanged. Site flooding and pipe break analysis are unaf f ected by the current method of positioning the anti-flood cross tie valves. Equipment inoperatility due to flooding, both internal and external to the reactor butiding. is evaluated. Since no new flood scenarios are created by the current methodology for positioning the anti-flood valves no new equipment malfunction scenarios are created, i
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Procedures. Tests & Emperiments Page 14 Safety Evaluation No.: 89-0104 bfSAR Section 6.2 & Table 6.2-8 2mplementation Document No.: Design Spectfication 3071-360 System No.: A3200 Title of Change: Information Addition to Design Specification 3071-360 Summary:
Articles 6.0 through 11.0 ' Underwater Destudging. Inspection and Repair of Torus Immersion and Air Space Areas." were added to Design Specification 3071-360 for maintenance of torus interior coating under submerged and wet conditions.
Saf ety Evaluation Summary:
Revision B to Specification 3071-360 provides the necessary instructions and additional coatings required to perform maintenance of the pressure boundary (suppression chamber) during submerged and wet conditions. These additional coatings and their application are compatible and in compliance with the same requirements (Reg. Guide 1.64 and ANSI 101.4) as the original coating system. The addition of coatings made by Revision B, when applied using the instructions provided, will be compatible with and function the same as the originat coating system..
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Implementation Document No.: DER 89-0726 i
$ystem No.: E1166 Titto of Changet Fan Motor Brake Design Evaluation Summary 4
This is an evaluation of the ultimate heat sink (VHS) mechanical draf t cooling tower (MDCT) fan motor brake design that is provided for tornado protection. This evaluation determines the suitability of the design as measured against the pertinent regulatory criteria, UFSAR commitments, and engineering design critoria. The answers to safety evaluation review questions are determined based on a comparison of the existing design to the final safety review analysis (FSAR) descriptions of the system upon which the NRC review and acceptance (Ref erence NUREG 0798) was based.
Safety Evaluation Summary:
7 The saf ety evaluation along with a probabilistic risk assessment (PRA) analysis shows high availability, despite the lack of single' failure proof design. (The design deficiency was corrected in 1989 under EDP 10518). A f ailure of the equipment could conceivably lead to missile generation, but the design has provided f or that and does not change the updated final safety analysis report (UFSAR) evaluation of missile damage. The brakes, though mounted seismically, can be postulated to activate following a seismic event due to relay contact failures. However, the relays are the same model and type that have been quellfied and the operating procedure chang, and operstor training will assure that the ultimate heat sink capability is not compromised. The PRA results show that the probability of equipment malfunction is et an acceptably low level, since the probability of f ailure of the UHS capability is an order of magnitude less than the generally accepted criteria of 10- events / year (Reference SRP 2.2.3 NUREO 0800).
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Safety Evaluation No.: 89-0131
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Implementation Document No.: NPP 20,205.01, Rev. 5 System No.: E1100 1
1 Titte of Change Procedure Change i
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Summary:
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Procedure NPP 20.205.01 was revised to include actions for loss of shutdown j
cooling in Mode 5 (Refueling) with water levet greater than 20 f eet 6 l
inches above the reactor vessel flange.
I Safety Evaluation Summary:
i This chanDe specifies the actions to be taken on the failure of the equipment and thus improves the capability of the operating staf f to respond to failures. The UFSAR indicates that att valve operations are by remote manual control. The proposed change only requires local manuel operation if a toss of SDC occurs. This does not degrade installed plant j
equipment resistance to failure.
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Safety Evaluation No,t 89-0133 i
Imptomentation Document No.: NPP 23.205. Rev. 30 i
System No.: E1100 Title of Change: System operating Procedure Change Summary:
A change was m t's to system operating procedure NPP 23.205, " Residual Heat Removat System, which allowed the residual heat removal (RHR) system to operate in a flowpath that allows cooling of fuel bundles located in the
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spent fust pool and the reactor vesset simultaneously.
Saf ety Evaluation Sunsaary:
Accidents that chattenge the f uel would be related to a positive reactivity addition (power increase) or a loss of coetant (heat removal capability).
This change to the UFSAR flowpath will provide adequate cooling in both the spent fuel pool and the reactor vessel. Positive reactivity additions due to shutdown cooling are an anatyred event. The initial condition shutdown cooling (SDC) equipment lineup as specified by Technical Specification limiting conditions for operation will be met by outage controls of plant equipment. The proposed split discharge flowpath heat removal capacity is adequate to remove the decay heat f rom both the spent f uel pool and reactor pressure veseet for att the conditions of loading /untoeding expected to occur durier the outage.
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procedures. Tests & Experiments Page 18 i
h Safety Evaluation No.: 49-0136 Implementation Document No.: DER 89-0896 System No.: R3000 Title of Change Emergency Diesel Generator Evaluation Summary:
Evaluation of the impact on operation of emergency diesel generator (EDG)
- 14 having one of the single phase Westinghouse Type CA dif f erentist current relays operating at a tower operating current than specified in the calibration procedure.
Safety Evaluation Summary:
The lower setting will not impact any equipment availability, i.e., the emergency diesel generator (EDG), as relay operation only occurs for a f ault in the generator and the lower setting only means a slightly lower fault current is needed to cause relay operations. Earlier fault detection may reduce the generator damage and repair time and costs. The relay only impacts EDQ operation and the smatt change in differential percentage for operation still has a large margin over the normal 0% difference current expected it. the generator.
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Safety Evaluation No.: 89-0143 UFSAR Section 5.2 Implementation Document No.: NE-5.5-181-NDE System No.: N/A Title of Change ISI Paogram Plan Changes Required by NUREG-0313. Rev. 2 and Generic Letter 88-01 Summary:
Changes were required to the ISI Program Plan and UFSAR as a result of NUREG-0313. Revision 2 and Generic Letter 88-01.
Engineering review of applicable austenitic stainless steel pine wolds, has resulted in classification of Fermi 2 welds into three of the categories identified in NUREO 0313 Rev 2, and Generic Letter 88-01.
A different percentage of weld inspections are required f or each category based on IGSCC crack susceptibility. All Category "C" welds required examination during the first Refuel Outage. Additionally, 26 ANSI B31.1 welds require inservice inspection due to classification as category "G" welds. As a result of the pending change to Technical Specification 4.0.6, a significantly larger population of welds will be rwquired to be inspected during all plant out s <,e s.
Safety Evaluation Samary:
Significantly more welds will be inspected to provide for early detection l
of IGSCC. This will allow more timely repair to reduce the chance of l
passive component failure thus reducing the probability for leakage from the reactor coolant system.
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l Safety Evaluations Procedures Tests & Experiments Page 20 Safety Evaluation No.: 89-0162
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Implementation Doctament No.: Work Request 003C890620 System No.: 82100 Title of Change Minimization of Intergranular Stress C ~ ~ision tracking i
Summary:
The implementation of the mechanical stress movement process (MSIP) is being performed to complete the mitigation actions for prevention or minimization of intergranular stress corrosion cracking (ICSCC) in the susceptible stainless steel reactor coolant pressure boundary (RCPB) piping and the reactor pressure vessel (RPV) nortles.
Safety Evaluation Summary:
The reactor pressure vessel (RPV) components, the reactor recirculation system components and the reactor water cleanup (RWCU) components treated by the mechanical stress improvement process (MSIP) aro passive components. Implementation of MSIP decreases the potential f or cracking in the piping being treated. The impiteentation of MSIP will not affect any ot or safety-related system er safety-related component which is required to have en active function.
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Safety Evaluations Procedures, Tests & Experiments Page 21 befety Evaluation No.: 89-0171 Figure Change Implementation Docuument No.: RERP Plan System No.: N/A Title of Change RERP Plan Revision 1 Summary:
All referentes to the State On-Scene Emergency Operating Center have been deleted from the RERP Plan. The State has revloed the Michigan Emergency Preparedness Plan to eliminate this facility. The functions will be carried out at other state facilities. This RERP Plan change resulted in a figure change in the UFSAR.
Safety Evaluation Summary:
The RERP Plan change that caused the UFSAR figure to be changed was reviewed in accordance with 10CFR60.54(q), and NRC was notified of the Plan change under separate cover (NRC-89-0217 dated October 20. 1989).
Safety Evaluations-Procedures, Tests & Experiments Page 22 Safety Evaluation No.: 89-0226 Implementation Document No.: NPP-24.202.01 System No.: E41 Titte of Changes HPCI Valve Survoittance Procedure Sumenary:
Valves F028 & F029 are the isolation valves in the 1-inch line f rom the HPCI drain pot to the main condenser. During a HPCI operability test the valves were not indicating full closed. The surveillance procedure was revised to remove the cperability of the HPCI valves f rom the acceptance criteria for HPCI operation.
Saf ety Evaluation Sunenary:
The valves are normally open to keep the HPCI steam supply line f ree of water. The valves close when the HPCI steam admission valve opens. The valves do not provide primary containmer.t isolation Operability of the HPCI system is not affected by the position of thoce valves. The amount of steam released through the 1" drain line is insignif t. ant compared to the amount of steam being supplied to the HPCI turbine. In addition, any release path created by these valvas in the open position is bounded by other steam release scenarios.
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Safety Evaluations Procedures. Tests & Experiments Page 23 UFSAR TEXT AND FIGURE CHANGES A number of text and figure changes were made to the UFSAR to reflect actual plant conditions. A safety evaluation was performed. These changes are summarized below.
Safety Evaluation No.: 88-0003 UFSAR Sections SA.1.3.3.f This change removed section 9A.1.3.3.f f rom the UFSAR. which described the f unctions of the fire protection committee. The description of the fire protection committee was incorporated into the UFSAR in WWech 1987. The comnittee is not part of the approvec fire protection program which is required to be maintained by the operating license. The fire protection committee provides a review function only of the program. Elimination of the committee has no impact on the level af fire protection in the plant.
Safety Evaluation No.: 88-0232 UFSAR Sections 11.4.3.8.2.15. Table 11.4-1 and Figures 11.4-2 & 11.4-5 The UFSAR was revised to delete ref erence to the Breathing Air Radiation Monitoring System. This system was originally designed to be used if breathing air was supplied by the Station Air System. Since we do not intend to use the Station Air System to provide breathing air, the Breathing Air Radiation Monitoring System does not need to be installed or operable. No credit is taken for the Breathing Air Radiation Monitoring System in UFSAR Chapter 15 Accident Analysis or the Technical Specifications.
Safety Evaluation No.: 89-0005 UFSAR Figures 5.5-6 & 5.5-8 through 5.5-12 Deleted ref erence to the RHR steam condensing mode which is not a part of the plant design.
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Safety Evaluation No.: 89-0007 i
I UFSAR Sections 9A.5 The UFSAR text was revised to change the description of the type of fire hose nogales used at the fire hose station outside the control room. The type actually used are adjustable pattern tog nozzles as opposed to fog only nozzles. Thit was a UFSAR text change only and was made tu provide consistency within section 9A.S.
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89-0008 UFSAR Figures 5.5-8 through 5.5-12 and Section 5.5.6.3g Deleted ref erence to the automatic open f eature f or RCIC valves E61F007 and E51F008, inboard and outboard steam supply line isolation. They are remote manual valves and close automatically on RCIC isolation signals. The automatic open feature is not part of the plant design.
Safety Evaluation No.: 89-0011 UFSAR Section 5.5.6.2.2.6 The UFSAR section was corrected to indicate the actual Condensate Storage Tank (CST) minimum storage tevet requirement is 150.000 gallons.
Safety Evaluation No.: 89-0025 UFSAR Figure 15.0-2 The change entailed on increase in the Fermi 2 cycle 1 end-of-cycle (EOC) integrated exposure ratio (INER). Cycle 1 was designed based on a maximum EOC INER of 1.0056.
This change evaluated increasing that to 1.0300, thereby creating an EOC axial exposure that is more bottom peaked than design. Operation with higher EOC INER affects both the axial power and exposure distribution; this does not affect plant conditions outside the core. The harder bottom burn will affect scram reactivity because the negative reactivity inserted during the first few feet of the core will be
f Safety Evaluations Procedures. Tests & Experiments Page 25 less since the f ust will be desder there. However, with the eight rods withdrawn no f urther than position 30, scram reactivity requirements are met and the transient events dependent on scram reactivity for mitigation (generator trip and f eedwater controller f ailure) have been calculated and shown to meet UFSAR requirements.
Safety Evaluation No.: 89-0055 UFSAR Section 11.3.3.6 The UFSAR section was revised to correctly describe the location of the Of f gas system radiation monitor as being at the discharge of the 2.2 minuto delay pipe.
Safety Evaluation No.: 89-0067 UFSAR Section 13.1. 13.5, 17.2 end Figure 13.1-2 & 13.1-3 The UFSAR was revised to show a change in the Nuclear Operations organization such that the Whintenance and Mooification department is separated. The Modification group will report to the Assistant to the Vice President Nuclear Operations and the Maint9 nance group reports to the Plant Manager. Maintenance is also reorganizing into 3 disciplines Electrical.
WWehanical and Instrument and Controls (IEC). The changes within the Whintonence organization have no ef f ect on the control of work in the plant.
Safety Evaluation No.: 89-0094 UFSAR Section 3.5.1.3 The UFSAR was revised to include tornado missile hazard analysis f or the reactor / auxiliary building exterior wall penetrations. HVAC intake enclosures and doors. The tornado missite hazard analysis concluded that the probability of missile damag6 to the vulnerable areas on the exterior walls of the Reactor / Auxiliary Building was acceptably low enough not to be considered a design basis accident.
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Procedures. Tests & Experiments Page 26 Safety Evaluation No.: 89-0097 UFSAR Sections 3.8.3.7 & 3.8.4.7 Two sections of the UFSAR were revised to clarif y the testing, inspection and surveillance requirements for structures, as they apply during the construction phase and the operating phase.
A*.so, the statements that routine and periodic inspections of concrete and structural steel members will be conducted were deleted. The structural elements affected by the changes are designed and constructed according to various industry codes as committed in the UFSAR. The allowable stresses in these codes are not contingent upon any in-service surveillance during the lif of the structure and thus, no suevoittance requirements are mandated by these codes. There are no environmental factors during the operation of nuclear power plant facilities which will affect the load carrying capacity of the structural elements under consideration. All the environmental parameters and seismic conditions which are expected during normal operation were accounted for in the design.
Safety Evaluation No.: 89-0101 UFSAR Section 13.1 R 17.2 and Figures 13.1-2 & 17.2-1
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The UFSAR was revised to show the Supervisor-Independent Safety Engineering Group (ISEG) reporting to the Vice President - Nuclear Engineering and Services rather than the Chairman, Nucione Saf ety Revis.* Group (NSRG). The Independent Safety Engineering Group (ISEC) provides an independent review of various aspects of plant operations as specified in the plant Technical Specifications.
Safety Evaluation No.: 89-0105 UFSAR Section 5.5.7.5 The UFSAR was changed to describe how the integrity of the RHR heat exchanger tubes is checked and ensured. The UFSAR currently states that a leak test is perf ormed on the tubes each ref ueling. This change reflects that tube leakage will be monitored on a monthly basis by monitoring the service water offluent radiation levels. This was not a physical change to the system.
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1 Safety Evaluation No.: 89-0109 j
UFSAR Section 6.3.2.2.3 and Table 3.9-27 UFSAR Table 3.9-27 was revised to state the correct design pressure and temperature of the core spray system injection piping between the inboard and outboard containment isolation valves. Also Section 6.3.2.2.3 was revised to clearly define the outboard valves by valve number.
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Safety Evaluation No.: 89-0145 UFSAR Section 6. 2.1. 2. 2.1
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The UFSAR was revised to clarif y the design b9ses f unction of the electrical interlocks on the secondary containment airlock doors. The interlocks are not required f or the airlocks to f unction and maintain secondary containment. They aid in the administrative control of assuring at least one airlock door is closed prior to opening the other.
Malfunction of the interlock circuits will not prevent the doors from perf orming their saf ety-related function of maintaining secondary 1
containment.
Safety Evaluation No.: 89-0146 f
UFSAR Section 13.1 and Figure 13.1-4 W
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l The UFSAR requirement that the Operations Support Engineer have a senior reactor operators (SRO) license was changed to require that the Operations Support Engineer either be SR0 certified or hold a SRO license. This position does not provide direct supervision of operation cf reactor j
controls. The Operations Support Engineer will only be eligible to fulf ult the duties of the Operations Engineer if licensed. The individual is still required to have completed the training necessary for an SRO (SRO certificate) and so has SRO levet knowledge.
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Safety Evaluation No.: 89-0151 UFSAR Sections 10.4.4. 15.2.2 & 16.2.3 and Table 15.2.3-1 This change was made to clarif y the UFSAR descriptions of the main turbine (active) bypass valves failure mode on a less of DC power. Although UFSAR l-
f Safety Evaluations Procedures Tests & Experiments page 28 Table 10.4-1, item 2, specifically states the bypass system is disabled on toss of 130 VD0 (B0p) battery potential, other sections of the UFSAR state
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a loss of the bypass system would require multiple random f ailures in the system. The system would require multiple f ailures to completely disable both bypass valves. However, the D0 power supply is not considered part of the system, and f atture of that power supply would disable the system.
Safety Evaluation No.: 89-0193 UFSAR Tabte 7.6-2 The UFSAR table was revised to indicate the corrac t fo' discharge temperature alarm setpoint. The UFSAR value (200 f) vid not agree with the G.E. design specification and plant procedure (220 F).
Saf ety Evaluation No. : 89-0195 UFSAR Figure 7.5-10 The figure was revised to make an editorial change "TAE" to "TAF" for the abbreviation " top of active fuet." TAF was deleted from the Reactor Level curve at t=50 min. since actual water levet is actually slightly above the TAF.
Finally, the suppression pool temperature curve was revised to indicate that a euppression pool temperature limit of 160 F is reached in t=76 min verses 200 min. as previously shown. The new curve is based on a plant specific calculation while the previous curve was generic.
The purpose of the suppression pool temperature curve is to show that the
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time to establish suppression pool cooling as described in UFSAR Section l
7.5.2.5.4 (60 minutes) is less then the predicted time to reach the temperature limit (76 minutes). This analysis shows that the time margin is reduced, however, an adequate amount of time exists to establish suppression poet cooling.
Safety Evaluation No.: 89-0204 UFSAR Section 4.2.4.3 This section was revised to describe Fermi's pre-established Faited Fuel Action Plan for detection, analysis, reporting and taking corrective action whenever fust failures occur. Eddy current testing was added to the list l
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e Safety Evaluations Procedures. Tests & Experiments Page 29 of inspection techniques. Finally, a discussion was added of Form 1's plans to participate in a vendor's lead fuel / test assembly program to obtain fuel performance data for en improved fuel design. This program is scheduled to commence with the second ref ueling outage and perf ormance inspection to commence at the third refueling outage.
Safety Evaluation No.: 89-0216 UFSAR Sections 3.6.2.2.1, 3.6.2.2.2
& 15.6 The analysis of the postulated main steam and f eedwater line break in the steam tunnel is described th UFSAR Section 3.6.2.
The analysis was completed in the early 1970's, and documentation supporting the conclusions reached could not be retrieved. Therefore a new analysis was done to support the UFSAR conclusions. Roanalysis of the main steam line break shows that the break flow is bounded by those described in the UFSAR 3.6.2.2.1.. and the off site release remains the same as previously evatusted in Chaptee 1E of the UFSAR. Flooding levels due to the main steam line break are boundea by the feedwater line break, and the resultant peak temperatures. pressure end humidity ef f ects on E0 equipment and plant structures have boon evaluated and found to be acceptable. Although flood levels are higher f or some areas of the p14nt. the consSquences of feedwater line break are bounded by tre analysis in UFSAR Section 3.A.4.4 Internal Flood Protection. The f eedwater line break of f site dose increason due to an increase in mcas flow. However, the results are ctill within 10CFR100 guidelines and are bounded by the main steam line break release.
Safety Evaluation No.: 89-0216 & 89-0098 Control Center Heating. Ventilating and Air Conditioning (CCtfVAC) and Standby Goa Treatment System (SGTS)
UFSAR Appendix A Conf oresnce to Regulatory Guide 1.52 SE 89-0098 was prepared to evaluate whether the existing configuration of the CCHVAC and SQTS duct work which was determined to be dif f erent than as described in the UFSAR constituted an unreviewed safety question. The evaluation concluded that the systems could perf orm their f unctions and an unreviewed safety question did not exist.
The following changes were made to Appendix A of the UFSAR text to reflect actual design, maintenance, and testing of the Control Center HVAC (CCHVAC)and SGTS ventilation systems.
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Regulatory Position 2.0.d for SGTS The maximum system pressure, corresponding to the SGTS f an peak pressure is l'
20.0 inches H 0 rather than 0.20 inches H 0 as presently listed in the 3
9 UFSAR. This Is a typographical error and does not impact the ability of i
the SGTS to perform as designed and evaluated in the UFSAR.
Regulatory Position 3.0.$ for CCHVAC & SGTS The UFSAR was changed to reflect that Fermi 2's charcoal absorbers do not use individual cetts, but are of a gaskettest design where the bulk absorbent material is filled and evacuated f rom the absorber bed by the use of special equipment (blowers, vacuums, etc). This type of design is also permitted by ANSI N509-1976 and has the advantages of being gasketless with less potential for bypass leakage, t
Regulatory Position 3.0.n for CCHVAC & SGTS This regulatory position requires the duct work to conform with 5.10 of ANS! N509-1976. The UFSAR was revised to state foe CCHVAC that the duct work ccaforms to the intent of ANSI N509-1980 for att areas of duct construction and testing. Under the worst loading combination of maximum internal pressure, plus dead weight and earthquake, some of the rectangular ducts will locally yield at the corners. Cross sectional flow area reductions of no more than 10% will result. No breach of pressure boundary i
integrity will occur.
For SGTS the duct work conf orms to the intent of ANSI N509-1980 for all areas of duct construction and testing.
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Regulatory Position 4.0.b for CCHVAC This regulatory position requires that system design provide a minimum separation of 3ft. betwoon banks of components in fitter units for ease of maintenance. Fermi 2's components are arranged such that the components can be replaced through access doors without having to have a person crawl into the unit, thus providing suf fielent space and access f or component replacements and maintenance.
Regulatory Positten 5.0.a for the CCHVAC and SGTS This regulatory position requires that a visual inspection in accordance with ANSI N510-1975 be performed prior to in-place testing of components.
The UFSAR incorrectly stated the regulatory position by listing the 1980 edition instead of 1975. However, Fermi 2 does comply with the latest edition (1980) of ANSI N510 as noted throughout the UFSAR. This is considered a typographical error and does not impact the ability of the SGTS and CCHVAC System to perf orm as designed and evaluated in the UFSAR.
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I Regulatory Position 5.0.c for CCHVAC and SGTS The UFSAR elso incorrectly listed the 1980 version of ANSI N510 f or in-place Diocryl Phthalate (DOP) testing of HEPA filters under this regulatory position.
Reg. Guide 1.62. Rev. 2 lists the 1976 edition. The two code editions are essentially identicat. Tte on\\y major difference being that the 1975 edition permits DOP testing of individual filters (or groups of filters) by the use of shrouds. rather than testing an entire bank of filters all at once. This type of test is not included in the 1980 edition and is not used at Fermi 2.
Regulatory position 6.0.d for CCHVA0 & SGTS A simiter situation ex*sta for the in-place testing of charcoal absorbers as with the HEPA filter DOP testing described above. The UFSAR listed the wrong edition of ANSI in the regulatory position. The two code editions are essentially identical. Except f or some editorial changes, the 1980
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edition permits the use of more types of ref rigerants as tracer gases.
Regulatory Position 6.0 on CCHVAC I
This regulatory position involves laboratory testing f ce activated carbon used in charcoat absorber beds. A clarification was added that states both
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the make-up and recirculation charcoal beds are treated separately and the l
laboratory testing of the carbon is done to acceptance criteria that allows l_
.a 95% gaseous iodine removat efficiency to be used in the dose i
calculations. This resu\\ts in a combined ef ficiency of 99.76% f or the j
makeup air because it gets filtered twice. This clarification is required i
because Fermi 2's configuration is different, in that a rscirculation flow l
1s introduced between the filter units creating a mixing situation prior to the recirc filter unit, resulting in a random re-fittering (double filtration) of makeup air.
Safety Evaluation No.t 89-0219 UFSAR The UFSAR was revised to incorporate certain fire protection Tech Spec Limiting Conditions for Operations (LCO) and Surveillance Requirements in accordance with Generic Letter 88-12.
The f ollowing Tech Spec Sections were added to the UFSAR Sections as indicated:
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Safety Evaluations procedures. Tests & Experiments Page 32 TECH SPEC UFSAR 3/4.3.7.9 FIRE DETECTION INSTR 4tENTATION 9A.6.1 3/4.7.7.1 FIRE SUPPRES$10N WATER SYSTEMS 9A.6.2 3/4.7.7.2 SPRAY AND/OR SPRINKLER SYSTEMS 9A.6.3 3/4.7.7.3 CC SYSTEMS 9A.6.4 3/4.7.7.4 HAf0NSYSTEMS 9A.6.6 3/4.7.7.6 FIRE HOSE STATIONS SA.6.6 3/4.7.7.6 YARD FIRE HYDRANTS AND ASSOCIATED 9A.6.7 HYDRANT HOSE HOUSES 3/4.7.8 FIRE RATED ASSEM0 LIES 9A.6.8 TABLE 7.1.1-1 APPENDIX R INSTRUMENTATION TABLE 9A.6.9.1 AND CONTROLS BASES RA.6.10 The administrative controte f or Fire Brigade staf fing as presently worded in Tech Spec Section 6.2 are streedy included in UFSAR Section 9A.1.3.3.
No change to this UFSAR Section was required.
The Tech Spec Section on Appendix R Alternate Shutdown will remain in Tech Specs per Generic Letter 86-12 and therefore was deleted from UFSAR 9A.6, Howevee, the list of Appendix R instrumentation and controls will remain in UFSAR Table 9A.6.9-i.
This change also deletes UFSAR Section GA 6.9.1 and
.2 and 9A.6.10 on Appendix R operability and surve111ance requirements which are retained in Tech Specs.
Safety Evaluation No.: 89-0220 UFSAR F1 purse 6.6-9 and 6.3-2 These figurop were revised to remove the mass.ftow rate vetus for HPCI and RCIC high steam line differential pressure trip setpoint and to correct the l
inches of water differentist pressure trip setpoint on UFSAR Figures 6.6-9 l
and 6.3-2 to be consistent with Design Calculations and Tech Spec Toble 1
3.3-2-2.
I Safety Evaluation No.
90-0013 UFSAR Sections 4.1 through 4.4 j
These UFSAR sections were revised to reflect the latest NRC appeoved version of Genera 6 Electric licensing report " General Electric Standard Application for Reactor Fuel" (GESTAR-II) NEDE-24011-P-A including the
" United States Supplement." NEDE-24011-P-US. Applicable sections of this
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Safety Evaluations Procedures. Tests & Experiments
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Pope 33 report are ref erenced and is consistent with the NRC overall standardization philosophy.
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The generic inf ormation contained in GESTAR-11 is supplemented by cycle-specific information and analytical results. The cycle-specific
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information includes a list of the fuel to be loaded in the core and safety I
analysis results. Cycle 1 information is documented in Chapters 4, 6, 6,
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and 16 of the UFSAR and subsequent cycle-specific inf ormation in Appendix B j
of tbo UFSAR.
GESTAR-!! is a report that documents the NRC approved GE safety analysis methodology and information specific to GE BWR plants in the United l
States. This methocology was used for licensing Fermi 2, Cycle 2 reload
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licensing approved by the NRC (Amendment 42 and 44).
The GESTAR-!! NRC
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l approved methodology provides the same margin of safety as the previous CE 1
approved methodology.
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l Safety Evaluation No.: 90-0016 UTSAR Sectione 1.1, 6.2.2.3, 6.3.3, 16.0, 16.1 and Appendix B l
The purpose of this revision was to support the reload fust portions of l
UFSAR Revision 3 which basically reflects the ef f ects of the Supplemental Reload Licensing Submittal for Reload 1 Cycle 2.
Most of the UFSAR revisions are contained in a new Appendix Bl the remainder are merely editorial changes in the base UFSAR and they ref erence the now Appendix B.
The licensing submittal has been reviewed and approved by the NRC.
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