NRC-92-0021, Safety Evaluation Summary Rept 1991

Safety Evaluation Summary Rept 1991

ML20090M262
Person / Time
Site:	Fermi
Issue date:	12/31/1991
From:	Orser W DETROIT EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-92-0021, CON-NRC-92-21 NUDOCS 9203240114
Download: ML20090M262 (154)

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References Fetui 2 4 , NRC Docket No. 50-341-NRC License lio. NFF-43

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+- :M th Annual 10CFR50.59 Safety E Evaluation Summary _Repor,t _,, _ _ _

a ant : to .10CFR50.71(e) and 10CFR50.59(b) (2). Detroit Edison her Leubmit . ision -5 .to1 the. Upda ted Final ' Safety Analysis Repo - FSAR) for: Fermi 2 a e annual Safety Evaluation Summary R 't .

The signed original and ten.

tional c.o -of the UFSAR. Revision 5.

are enclosed :one copy will be su to Region III and one copy to

-the NRC = Resident Inspector. . 'SAR co a changes made since submittaliof Revision-4 arch 1991'and as a m m describes the plant configuratl tough September 20, 1991, aix mon rior to this submitt al, nges- associated with Revision 5 are annotated vision

.:bara e appropriate margin marked with a_ "5". ~ All revised pages ed_Rev 5.3/92..

' Also enclosed: is the. annual -Safety Evaluation Summary Report containing a brief description of changes to plant dr ign, procedures, tests.

experiments and the UFSAR.

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USNRC-March 18. 1992 NRC-92-0021 ,

Page-2 I_f you hav any questions, please contact Evelyn F. Madsen at G13) 586-4205.

Since rely.

Enclosure AU/

cct T. G. Colburn A. B. Davis R. W. DeFayette (one copy of UFSAR)

S. Stasek (one copy of UFSAR) 5 4

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USNRC March 18, 1992 NRC-92-0021 Page 3 I, W. S. Orser, do hereby af f irm that the foregoing statteents are based on facts and circutretances which are t rue and accurate to the best of my knowledge and belief.

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W. S. Orser Senior Vice President On this _

/ day of fj d.[ .__, 1992, before me personally apper *d W.____S. Orser, being first duly sworn.and says

- t hat he executed _he foregoing as his free act and deed.

/h'f4A$sv UXdWG j Notary Publie l- .-

i' Awua A AkMiri A I NUTARY P15UC STATE CT ISCIECM' Mor(AOE CLE'NTY My COMMstCN F_XP NCfGOM_

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U.S. NUCLEAR REGULATORY COMMISSIOt4 e4C<OSts!"A"n%=

Fermi 2 SAFETY EVALUATION

SUMMARY

REPORT 1991 Detroi~:

EDISON

Enclosure to NRC-92-0021 FERMI 2 SAFETY EVALUATION

SUMMARY

REPORT 1991 Docket No. 50-341 License No. NPF-43

FERMI 2 SAFETY EVALUATION

SUMMARY

REPORT 1991 AS-BUILT NOTICES 1

MFETY EVALUATIONS AS-BUILT NOTICES Page 1 SAFETY El ALUATION GLMMJtY Safety Evaluation No: 88-0052 UFSAR Revision No. 5 Sectlon(s) 6.4; 9.4 Reference Document: _A9N 8660-1 Table (e) N/A Figure Change (X) Yes [ ] No Title of Change: Control Center Heating, Ventilation. and Air s <ttioning (CCHVAC) System Setpoints SLA&MRY:

This as-built notice revised the master instrument list and CCHVAC mechanical system diagrams to reflect the as-built control setpoint for the normet and recirculation modes of the CCHVAC system (+ 1/4" W.C.). This control setpoint for both modes of CCHVAC in consistent with the control conter pressure requirement (+ 1/4" + 1/8" W.C.) provided in the UFSAR.

This as-built notice revised the master instrument list and CCHVAC drawings to reflect the as-built control setpoints and control pressures identified in the UFSAR. Consequently, this is a documentation change only. The context of this notice does not impact the operation of CCHVAC in the recirculation mode.

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SAFE 7V EVALUATICNS AS-BUILT NOTICES Page 2 SAFETY EVALUATION SLMAARY Safety Evaluation No 88-0232 UFSAR Revision No. 5 Reference Document: ABN 9877-1 Sectlon(s) N/A Table (s) N/A ligure Change (X) Yes [ ] No Title of Change: Breathing Air Radiation Monitor System Changes S1M MRY:

Various sections of the UFSAR describe the breathing air radiation monitor system. This system would have been . used if station nie was utilized for breathing air purposes. Since there is no intention to use station air for breathing air, the breathing air radiation monitors will not be installed and nade operational.

These radiation monitors are not subject to any equipment malfunctions considered in the UFSAR, Further. there is no process or electrical link between these monitors and other plant equipment. Regulatory Guide 8.15 for monitoring station. air radioactivity has not been violated as station air will not be used as breathing air.

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SAFETY EVALUATIONS AS-BUILT NOTICES Page 3 SAFETY EVALUATION SLAMARY Safety Evaluation No: 90-0162 UFSAR Revision No. ,

5 Reference Document: ABN 12059-1 Section(s) 9A.42 9A.5 Table (s) N/A Figure Change IX) Yes I 1 No Title of Change: Reactor, Auxiliary, and Turbine Building Walt Fire Barrier Upgrade GLASMRY:

This evaluation justifies upgrading the following exterior p'. ant walls to rated fire barriers:

1. The north, south, and west exterior walls of the reactor building below the metal siding at elevation 684'-6".

2. The north and south exterior walls of the auxiliary building including the cable vault roof located on the south wall.

3. The west exterior wall of the turbine building below the metal siding at elevation 679'-6".

These walls are being fire rated to ensure that trailers and other combustible materials that are located in close proximity to tne plant during refuoting outages and for other possible reasons do not present a-fire hazard to safety related equipment and circuits within the plant, f The above walls are at 1,e a s t 18" thick. Underwriters Laboratories, the National Fire protection Association, and the American Concrete Institute l confirm that a reinf orced concrete wau with a minimum thickness of 8" is a 1 3-hour rated fire barrier. All penetrations and openings in the above walls i are either 3-hour fire rated penetration seals or are technicauy justified as l . adequate fire stops in accordance with the guidance in NRC Generic Letter 86-10. The original untabeled cable vault security door has been replaced with a 3-hour rated door. This door continues to meet security plan requirements, i

The previously approved Appendix R analysis assumed that combustibles are maintained at an acceptable distance from safety related buildings and that no fire in the yard area would cause damage or enter any of the fire zones facing

, the unrated exterior walls. Providing a rated fire barrier in an exterior wall is an acceptable alternative suggested by NRC Generic Letter 86-10 (Reference Questions and Answers 3.1.3.) and, therefore, assures that this assumption is stil.1 valid.

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SAFEVY EVALUATIONS AS-O' JILT NOTICES Page 4 SMLIY LVALUAI1ON SUWARY Safety Evaluation No: 91-0054 UFSAR Revision No. 5 Reforence Document: ABN 11273-1 Sectionis) N/A Tabte(s) 6.2-2: 7.5-5: 9A.6.1-1 Figure Change IX1 Yes t 1 No Title of Change: Emergency Equipment Cooling Water (EECW) Drywell Supply and Return Valve Renumbering StM MRY:

This as-built notice-documents the interchange of valve number designations for Division I EECW drywell supply isolation valve P4400F606A and Division I EECW drywett return isolation valve P4400F607A. This change ensures that the valve

. numbering scheme for both the Division I and Division II valves is consistent.

License Amendment 70 incorporating the valve number change was approved by the NRC.

This change does not make any hardware changes. The evaluation of the hardware changes associated with the valve renumbering is documented in safety evaluation 90-0134 as part of engineering design package EDP 11273. The as-built documentation of this change will, img rove the man / machine interface by standardizing the numbering designations of the valves between Divisions I and II, 1 -- _ _ _ _ _ - _ . - - . _ . . . _ . . . . _ . . . _ . . - = _ _ _ _ __ - ...............

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Safety Evaluation No: 91 0055 UFSAR Revision No. 5 Reference Document: ABN 12531-1 Section(s) _N/A Table (s) 3.2-1 Figure Change IX3 Yes I 1 No Title of Change: Emergency Diesel, Generator (EDG) Standby Fuel Oil Pump Discharge Line Material Code Change SUhSAARY:

This safety evaluation justifies changing the specification code for each EDG ,

standby fuel oil pump discharge line from ASME III to the Diesel Engine Manufacturers Association (DEMA) code. The SB-75 coppar tubing and 58-164 fittings specified under tbc ASME III code are ns longer available. The diesel manufacturer recommends S-75 copper tubing and ASTM A276 type 316 fittings specified by the OEMA code as equivalent replacement parts. Reclassification of the parts from A*YE III to DEMA al. lows the use of market available copper tubing and fittings.

This specification change does not change the operation or function of the

• EDGs. The new parts wilt transfer fuel and mainteln the pressure boundary in the same manner as the original parts. The discharge line is still e QA 1 seismic category I instattation meeting all the regulatory requirements for the EDCs.

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SAFETY EVQLUAVIONS AS-BUILT NOTICES Page 6 SAFETY EVALUATION SLA9AMY Safety Evaluation No 91-0056 UFSAR Revision No. 5 _ _ _ _

Reference Document: ABN 8264-1 Section(s) N/A Tatde( s) N/A Figure Change (X1 Yes [ ] No Title of Change: Documentation of the Permanent InstaM ation of the Vibration Velocity Transducers for the Reactor Recirculation Pumps StA44ARY:

ABN 8264-1 documents the permanent as-built configuration of the Bentley-Nevada velocity transducers originauy instaned by temporary modification 85-057.

Three velocity transducers are installed on the frames of each pump. The transducers are used as a backup to the Robertshaw vibra-switch alarm switches. If a spurious alarm develops and the drywe n is inaccessible to allow raising the setpoint. periodic readings of the velocity transducers are made to ensure the velocities are less than 0.4 in./sec.

This modification does not change the function of the reci rcul.at ion pump vibration and monitoring system. The control circuits for this system are independent of the recirculation pump motor controls. As a result, this modification does not affect the operation of the recirculation pumps. A malfunction of the vibration monitoring and surveillance system does not contribute to the malfunction of other safety related equipment.

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SAFETY EVALUATIONS AS-BUILT NOTICES Page 7 SAFETY EVALUATION Slfmaay Safety Evaluation No 91-0060 UFSAR Revision No. 5 Reference Document: ABN 12319-1 Section(s) N/A Tab k(o) , N/A Figure Change (X) Yes ( 1 No Title of Change: Reactor Core Isolation Cooling (RCIC) System Diagram Revision

'NY:

This evaluation justifies revising the RCIC system UFSAR figure to include RCIC vibration monitoring equipment installed during pre-operational and startup testing of the RCIC turbine and pump. The equipment is not currently identified on any plant base configuration design documents, This equipment consisti of:

1 Ten vibration transducers mounted on the RCIC tarbine and pump bearing housangs.

2 Cne terminal box.

3. Ten interconnecting cables routed from the vibration transducers to the terminal box.

This equipment has been 1. eft in place due to poor accessibility. It will, remain permanently instatted and is utilazed for intermittent vibration monitoring in conjunction with portable vioration testing equipment provided by test personnel.

The vibration equipment is non-0,'its instattation has been field verifi6d, and it - conf orms to seismic II/I criteria. This equipment does not affect the

!. operation. function, or performance of any plant equipment. This modification does not introduce any new failure modes or scenarios that affect the reactor coolant or containment bnundaries.

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SAFETV EVALUATIONS AS-DUILT NOTICES Page 8

?AFETY EVALUATION StM.MRY Safety Evaluation No 91 0081 UFSAR Revision No. 5 Reference Document: ADN 12430-1 Section(s) N/A Table (s) N/A Figure Change 1X] Ves ( 1 No Title of Changer Reactor Recirculation Pump (RRP) Speed Limiter Setpoint Change SUheMRY:

This modification enanged the s2 (tripped feedwater pump) and #3 ( 1,o s s of heater drains) RRP speed limiter setpoints from 42% and 40%, respectively, to 37%. Th'.. was done to ensure that plant operation in the maximum extended operating domain (ME00) does not place the plant in the thermat hydrautic instability region on the power / flow map or result in a scram following the loss of a single feedwater pump or heater drains pump. The 37% setpoint corresponds to 48% core flow. This provides a 3% core flow margin to the instability region. Thir+ setpoint also allows the condensate /feedwater flow to match the capacity of the polishing dominerati2ers.

A transient- analysis was performed to determine the feedwater capacity available following a single feedwater pump trip concurrent with the loss of heater drains and the subsequent RRP runback. The results of thie essessment showed that there is a possibility of a reactor water low tevet scram following the runback from the upper and of the maximum extended 1,oad 1,ine 1,imi t (MELLL) region (75% core flow at 100% power). To avoid a tow water levet scram, the plant operating procedures require the operators to start the standby feedwater system (SBFW) when the finat reactor power is greater than 70%. The use of the S8FW system will help to maintain reactor water. level allowing a controlled rod insertion to reduce power to within the licensing domain, l These changes enhance the plant's capability to operate outside of the thermal hydraulic instability region'and to avoid a reactor scram, The recirculation system runback with the subsequent SSFW actuation is not an event analyzed in the UFSAR. However, it is bounded by the analyses for loss of feedwater flow, high pressure coolant injection, and the recirculation flow control failure transients. The function of the recirculation spend control 1,imi t s is not altered. The use of the S8FW system to mitigate the possibility of a reactor low water level scram is in accordance with the design objectives in the UFSAR

( and approved plant procedures.

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l SAFETY CVA.UAVIONS AS-BUILT NOTICES Page 9 i l

SAFETY ENALUATION $lDAMRY Safety Evaluation No: 91 0096 UFSAN Rev6sion ko, _

5 Reference Docapent; ADN 12649-t tactiLn(s) 9 7 ; A.1.

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latae ( s) N/A a-Figurc Cnengs t ) Yee (XJ Ho a

( Title of Changer Revising the Residuet (RHR) Heat Remcval, System Reservoir l Volume in the UFSAR SlMMRY:

1 This evstuation justified revising the UFSAR to correct the stated RHR l reservoir volumes. A revised design ca \.cula t ion was performed to ustermine the RHR reservoir volume between otovations 569 ft end 590 ft and reflects the as-built volume of the RHR reservoir. This revised calculation used more accurato dimensions from architectural and civit - drawings. The applicable results have been incorporated into the UFSAR to accurately describe the RHR reservoir volume for a given elevation. In addition, the UFSAR has been revised to reflect the correct remaining RHR reservoir volume after the 30 day LOCA supp.y is used. The figures in the design calculation were used to calculate this margin.

This revision does not change the design criteria, function, or operation of

l. the RHR system. The revir. ion to the desinn calculation is based on previously approved documents. The calculated RHR reservoir volumes stilt meet the minimum RHR reservoir 30-day water supply required by the Technicat Specification 3.7.1.5.

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SAFETY EVALUATIONS AS-BUILT NOTICES page 10 a e

MINOR ADN'S The fonowing As-Sull,t Notices (ABNs) resulted in UFSAR drawing or text changes. Tnese chonges were reviewed for potential safety consequences. Because the changes were minor and were made to reflect ea-buitt plant conditione. a surnmary f or each was not. prepared. The ABNs and their associated safety evaluations have been listed for

-reference.

Safety Evaluation No.: 88-0121 Figure Change Implementation Docunent: ADN 6815-1 Safety Evaluation No.: 91-0002 Figure Change Implementation Document: ADN 11662-1 Safety Evaluation No.: 91-0008 Figure Change Imptomentation Document: ABN 11982-1 Safety Evetuation No.: 91-0022 -Figure Change Implementation Document: ABN 11560-1 Safety Evaluation No.: . 91-0028 Figure Change Implementation Document: ABN 11854-1 Safety Evaluation No.: 91-0036 Figure Change Implementation Docunent: ABN 12163-1 Safety Evaluation No.: 91-0049 Figure Chang,a '

Implementation Document:' ABN 11645-1 i

l l- Safety Evaluation No.: 91-0068 Figure Change

! Implementation Document:. ADN 12345-1 L

Safety Evaluation No. .

91-0071 Figure Change Imptomentation Document: ABN 12186-1 Safety _ Evaluation No.: 0126 Figure Change Implementatson Document: ADN 12784-1

-END OF ABN SECTION 1

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FERMI 2 ,

SAFETY EVALUATION

SUMMARY

REPORT 1991 POTENTIAL DESIGN CHANGES 9

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POTENTIAL DESIGN CHANGES Page 1 SAFETY EVALUATION SLANARY Safety Evaluation Not 88-0119 UFSAR Revision No. 5 Reference Document: PDC 9090 Sectlon(e) N/A Table (a) N/A Figure Change (X) Yes I ) No Title of Change: North and South Reactor Feed pump Drip Drain Isolation Valves Removat and Piping Reroute SLAGAARY:

1his modification removed the north and south reactor feed pump drip jselotion valves and rerouted the south reacter feed pump drip drain piping. The isolation valves were removed because the possibility exists that if the valves were closed, the drip cavity would fitt with seat water. It is then possible for the water to enter the bearing oil area and contaminate the tubricating oil. The associated drip drain piping of the south reactor feed pump was rerouted from an equipment drain to the floor drain system. This was done bec.suse-the equipment drain system is not equipped to handle potentially city water. The floor drain system utilizes an cit / water separator before the water is processed in the redweste system. The design conditions for the drip drain piping were revised from 950 psig 9 430 F to 14.7 psia 9 212 F since the piping and the pump bearing hub are open to the atmosphere.

This modification insures the reliable operation of the reactor feed pumps by decesasing the- probability of water c,antaminating the tubricating oil.

R* routing of the drip drain piping decreases the potential of oil entering the equipment drain system. thus improving the operation of the redwaste system.

The subject piping and valves are not covered by the technical specifications.

Therefore, there is no impact on the technical specifications.

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SAFE?Y EVALUATIONS POTENTIAL DESIGN CHANGES Page 2 SAFETY EVALUATION SLMAMY Safety Evaluation Not 90-0151 UTSAR Revision No. 5 Reference Document: PDC 11455 Section(s) 4.11 4.2: 4.5 Table (s) N/A Figure Ctunge (X) Yes ( ) No Title of Change: Control Rod Replacement SlMMRY:

165 of the original 185 control rods were replaced during the second refueling outage. 96 control rods installed in the non-control cell positaons are Generat Electric Ouratife 140-C matched-worth rods and the remaining 69 replacement contret rods installed in the control cell positions are General Etsctric Duratife 215-C matched-worth reds.

The Duratife 140-C control rode beve the following changes:

1. The top handle, wing sheaths, tie rod, and bottom coupling segment are fusion welded.

2. Thicker wing sheaths are used and the internal stiffening strips have been eliminated.

3. The number of BC absorber tubes has been increased from 76 to 84 4

and the tubes have been shortened from 143" to 137".

4. A 6" hattnium plate is inserted in the top of each wing.

5. An extra row of cooling holes has been added at the top and bottom of each wing.

6. Low cobalt matarials are uced for the pins and rotters.

The Duratife 140-C control rods are approximately 10% heavier than the original control rods due to the higher density of the halfnium.

SAFETY EVALUATIONS POTENTIAL DESIGN CHANGES Page 3 Safety Evaluation No. 90-0151 (continued):

The Duratife 215-C Controt rods have the same changes as the Duratife 140-C control rods with the following excsptions:

1, The total number of BC tubes is reduced from 76 to 72. The inner diameter of the tubes has been increased to accommodate more B C per tube.

2. A baltnium metal edge strip runs the full 143' absorber length and hal.fnium plates are installed in the upper 6" of each control rod wing.

3. The bottom controt rod drive (CRO) coupling segment and integral velocity limiter have been redesigned to reduce overall weight such that this model controt red is essentially the same weight as the original model control rod.

The improved construction materials enhance control rod integrity and the halfnium extends service life.

The replacement rods are dimensionally compatible with the control rod drives and the reactor core confiruration such that there is no significant impact on controt rod scram times. The channet bowing tolerance. seismic loading limits, mechanical timits and material compatibility of the new controt rods are equal to or better than the original controt rods. The replacement controt rods are matched-worth and bounded by the existing lattice physics analysis and the accident and transient analyses. The minimum critical power ratio (MCPR) remains greater than the safety limit MCPR in the applicable analyses. The use of fusion welds as opposed to the overlay and spot welded assemoty of the original controt rods, improves the control rod mechanical integrity and forms a crevice frew structure which eliminates t he potential for crevice corrosion damage. NRC approved 1.icense Amendmer.t 66 anows the use of hal f nium at an absorber. The use of tow cobalt materints in the controt rod pine and rollars eliminates additional cobalt from entering the reactor assembly. The additional weight of the Duralife 140-C centrot rods causes a smatt increase in control cod scram times. However, the weight is within the design margin for the control rod drives and the overall negative reactivity insertion rate with the replacement control rods is within the accident analyses and technical specification timits. Haltnium hydriding incidents reported in NRC Information Notice No. 89-31 are not a concern at Fermi 2 because the replacement control r uds use bare hattnium exposed to reactor cool, ant flow. This prevents hydrogen dd f usion f rom occurring because hydrogen cannot concentrate and the pr:,tective half nium Wrf ace oxide 1 ays.r is maintained.

END OF POC GECTION

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l FERMI 2 SAFETY EVALUATION

SUMMARY

REPORT 1991 ENGINEERING DESIGN PACKAGES

SAFETV EVALUATIONS ENGINEERING CESIGN PACKAGES Page 1 MFETY EVA1.UATION SLWRY Safety Evaluation No: 07 0009 UFSAR Revision No. N/A Referunce Do m ont: EDP 4271 Sectlon(s) N/A Table (e) N/A Figure Change [ ] Yes [X1 No Title of Change: Instattation of a Maintenance Power Distribution System St M RY:

This modification ' installed a dedicated balance of plant (DOP) 480VAC maintenance powee distribution system in the reactor building. The system consists of . two main distribution panels feeding tocally mounted disconnect receptacles located throughout the reactor building.

The maintenance power distribution system is dedicated to supplying power to maintenanca loads and is powered from a balance of plant powar source that originates at Formt 1. This power system is not associatee with any plant process system (safety related or BOP). The additional first floor reactor buildir; fire loading contributed by the cable trays installed by this modification has already been accounted for in the fire hazards analysis.

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SAFETY EVALUATIDNS ENQ!NEERING DES!CN PACKAGES Page 2 SAFETY EVALUATION SLWWRY l

Safety Evaluation No: 87 0063 REV 1 UFSAR Revision No. _

5 Reference Document: EDP 6641 Section(s) 1.2; 3.1; 7.1; 7.6; 7.7 Table (s) 1.6-1; 7.7 1 Figure Chmgo (XJ Yes ( ) No Title of Change: Rod Worth Minimizer System Rep 1,acement SL9nM,RY:

This modification replaced the original rod worth minimizer (RWM) system with a stand alone microcomputer-based NUMAC RNA system. This changeout constitutes corrective actions to satisfy an NRC commitment to prevent recurrence of control rod manipulation errors and is part of the Reactor Operations Improvement Program.

t l The new RWJ design does not al,ter the purpose or function of the original RfM l

system. The reliability of the NUMAC RMJ is anhanced by its self-test capabilities and availability of spare parts. The electrotuminescent display

! provides the operator with a visual control irtterf ace that enhances human

! factors considerations. Like the previous RNA system, fatture of the new R)W system results in an automatic loss of rod movement permissives. The consequences of the RiW fatture are bounded by the control rod deep accident l

analyzed in the UFSAR.

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SAFEVY EVALUATIONS ENGINEERING DESIGN PACKAGES Page 3 SAFETY EVALUATION SLMMRY Safety Evaluation No: 87-0127 REV 1 OfSAR Revision No. 5 Reference Document: EDP 2189 Section(s) N/A Table (s) N/A Figure Change [X) Yes ( ) No Title of Change Reactor Building Exhaust Plenum Radiation Monitoe 011-P200 Sample Line Modifications SLANARY:

This modification sloped the sample lines down f-om the higher elevations; heat traced the sample tinos and sampler SA-13: added a condensate cettection bottle; and added bypass valving for additional sample points. This modification was-implemented to prevent condensation from f orming within the sample lines and possibly damaging radiation monitor 011-P280. This modification also allows maintenance and troubleshooting to be performed on the monitor without disconnecting the sample lines and anows t ernporary grab samples to be taken whe. the radiation monitor is inoperable.

This change is a hardware and system enhancement. AM hardware ins t aned . by this modification conforms to seismic II/I criteria. This change has no effect on any analyzed accidents in UFSAR chapter 15. There is no change to offsite or onsite radiation doses and there is no change to radioactive materlat releases.

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SAFETV EVAttmTIONS ENGINEERING DES! Odd PACKAGES Page 4 SAFETY EVALUATION SLM4ARY Safety Evaluation No: 89-0065 UFMR Revision No. 5 Reference Document: EDP 9734 Section(s) -N/A Table (s) N/A Figure Change (X) Yes i ) No Title of Changer Fuel Channet Heist Modification f

SLMERY:

The channet hoisting device located over the new fuel inspection stand was modified to facilitate channet handling activities. The instatted channot hoisting device was a temporary installation which was used during the initial fust channeling activities. This scheme was based on a GE proposal. The winch and motor assembly were located on the parapet watt and the rigging cable routed up to the roof levet and across just above the new fust inspection stand. This modification retains the same configuration. Therefore, the motor / winch assembly and their support are the same. This modification improved the structural components at the roof level with a new beam and the selection of proper rigging hardware.

This hoist cannot lift heavy loads and cannot move loads over the spent fust pool or open reactor vesset. The hoist's movement is in the vertical or near vertical direction. Its usage is limited to new channet instattation and maintenance activities. This modification was designed to perf orm saf sty for the rated toad capacity of 500-lbs. The magnitude of the load is smatt compared to the loads evaluated in " Load Drop Analysis of Heavy Loads".

Therefore. a load drop from this hoist will not damage the refueling floor stab structure. If the channel is dropped on the new fuel bundle, it may damage the fust but it will not have any radiological consequences. _This is not a new scenario since the new fust channeling activity has not changed. -A variety of events t ha t qualify as fust-handling accidents have been investigated. The most severe accident ts dropping a spent fuel bundla into the reactor core.

This accident produces the largest number of failed fuel rods. Therefore, the use of the hoisting device will not create the possibility of an accident of a difforent type than any previously evaluated in the UFSAR.

SAFETY EVAltlA?!ONS ENGINEERING DESIGN PACKAGES Page 5 SAFETY EVAlt1ATION SLMAARY Safety Evaluation No 89-0068 UFSAR Revision No. 5 Reference Document: EDP 9134 Section(s) 12.2 Table (s) N/A figure Ctange ( 1 Yes (X) No Title of Change: " Machine Shop pressure High" Alarm Window, Recorder point, and Local Horn Alarm Removat StsNARY:

This modifiestion removed the machine shop high pressure control center alarm window 8020, the associated sequentist recorder point 001X41, and the associated locat horn alarm. Originally, the machine shop was to function as a hot shoo. This required that the machine shop be maintained at a negative pressure with respect to the Office Service Building (OSS) to prevent the spread of contc.mination and the ex'itt'ation of contaminated air, Alarm window 8020 was provided to inf orm the operators when the pressure difierential between the OSB and the machine shop exceeded 1/8" W,C. Since the machine shop is not used as a he,t shop, the spread of contamination and the exfiltration of contaminated air is no longer a concern, Therefore, the alarm window, sequentist recorder point, and locat alarm horn are not required.

The equiptnent removed by this modification is not remired because the machine shop, by procedure, cannot be used as a hot shop unless special radiological controls have been established and approved by Radiation Protection. The machine shop has - an exposure level of less than 0.5me/hr and it is monitored for radiation release through an area radiation monitor. The function of the HVAC remains the same.

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SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES Page 6 SAFETY EVALUATION SUMMRY Safety Evaluation No: 89-0076 REV 1 UFSAR Revision No. 5 Reference Document: EDp (1740 Section(s) 3.10 Tabte(s) 3.10-3 Figure Change IX) Yes i 1 No Title of Change Reactor Pressure and Level Instrumentation Rack Replacement StM%RY s This modification replaced two existing reactor pressure and levet instrument racks with a completely redesigned version of each rack. This method of replacing an existing rack with a completely new rack was selected primarily to meet schedule requirements and Technical Specification operability restraints.

The time period tnat the individual or collective set of instruments installed on the racks can be out-of-service is limited to the period of time when irradiated fust is not handled in the secondary containment; there are no core atterations in progress; and there are no operations in progress with a potential for draining ine reactor vessel.

Instattation of the instruments and racks wilt not result in the increase in offsite or onsite radiological releases since the racks and instruments are I

designed to meet or exceed all applicable Edison. and NRC Regulatory I requirements. The enhanced human factors characteristics of the rack design i result an a decrease in the probability of a mal f unction due to surveillance testing. In addition, the single failure capability of the protective system is retained by this change.

SAFET" EVALUATIONS ENGINEERING DESIGN PACKAGES Page 7 SAFETY EVALUATION SlMAARY Saf ety Fivduation Not B9-0093 UFFAR Revision No. 6 Reference Documentt EDP 4274 Section(s.) 3.12: 8.3 Table (s) h/A Figure Change i ) Yes (X) No Titte of Change: Recirculation Motor Generator (MG) Set Control Panet Modifications SUWMRY:

Both Division 1 and Division 2 control cables enter the MG set control panels.

These cables are part of the recirculation pump trip (RPT) portion of the anticipated transient without scram (ATN ) system. This change provided additional separation between opposite division cables and wiring inside the control panels. This was achieved by the addition of physical barriers surrounding the cable of one of the divisions in each panol and providing a fire retardant barrier between the redundant trip coils (' C) on the generator field breaker to which the divisional cables are routed, The addition of fire retardant barriers internat to the MG set control panels enhances the ability of the systems to perform their intended safety function.

These barriers provide adequate protection for at least one of the QA.1 divisional trip signal circuits for any of the following hazards:

1. Short circuit of CA-1 circuit

2. Short circuit of a balance of plant circuit

3. Gross failure of one trip coit 4 Crimping of wire

5. Fuse failure

6. Small fire internal to panel The only adverse affect would be an insignificant addition to the combustible loading in the event of a major fire. However, these barriers are not intended to protect against a major fire nor is the equipment required for the Appendix R Fire Scenario.

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SAFE 7Y EVALUATIONS ENGINEERING DESIGN PACKAGES Page 8 SAFETY EVALUATION tRMMRY hafety Evduation No 89 0100 UFSAR Revision No. 5 Refeeonce Document: EDP 9828 f'.ec t i on t a ) N/A _

Table (e) . _ _

N/A Figure Change (X) Yes [ ] No Titte of Change Iso. Mimic Penel Modifications Suhta$.RY:

This modification removed the dynamic portion of the iso. mimic pane t, and replaced it with a status indication display showing the status of the main steam isolation valve (MS!V) trip togic circuits. The new display shows the status of the MSIV isolation relays plus sovon variables in_ each trip system that can cause a MSIV isolation f or each of the four channels. The variables being displayed for each channet are main steam line low pressure; reactor water low levet 1; reactor building and steam tunnet high temperature; main steam tine high flow; main steam \ine h19h radiations condenser tow vacuum and turbine buildireg high temperature. This information will help prevent ope'ator errors in assessing MSIV trip togic status.

Thic modification effects contros room annunciation. It does not affect a.:tomatic or manual operation of the MS!V trip logic cireut t a. There is no enange to the oneste or offsite radiation doses or radioactive tratoriet celeases. This modification installed an operator aid which functions in a manner similar to the esisting annunciation system. This modification does not affect the operation of any safety.retated equipment. A u new conduits instatted by this modification are seismically mounted in accordance with Fermi 2 standard specifications. Att new balance-of-plant (BOP) cablob are routed in DCP cable trays and conduits unt1\ they reach safety.related cabinets. In safety.related cabinets, BOP cable aru .aing is separated from cabinet internal wiring carrying safety.cetatet u gn$ as much as possible. Att new inputs are taken from dry contacts ei evit. 'ng relays. The status of each contact is seS *1 with a low energy y' source which cannot degrade the function of 1- safety-related circuits on the relay coils.

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SAFETY LVAltJATION SLL94ARY Safety Evaluation No: 89-011G REV 1 UFSAR Revision No. b Re f erence Docuhant s LOP 3303 Section(s) N/A Tabte(s) y/A __

Figure Change (X) Yes ( ) No titte of Changes Post Accident $smpting Systein (PASS) Ventilation System SUWARY:

This modification instetted a fan and filter assembly to provide ventitation at-the PASS sampling station. The primary purpose of the venti \stion system 19 to remove heat from the PASS sempting station to extend the life of electrical components. A ventitation duct is attached to the top of the PASS panet. Air is drawn through the enmpting station by a ventitation fan located outside the post accident sampting room. The fan discharges through a profitter, HEPA fittee, and charcoat adsorber into the \abyrinth area located touth of the post accident sampling room. Filtered air is .ubsequently drawn from the plant by the turbine building exhaust system. The ventitation fan will be operated whenever the PASS sampling station is operated. Manual fan control and indication to provided at a \ocat instrument rock. Fitter difforential pressure indication to available to the operatoa inside the post accident sampling room. Additionally, this modification replaces e temporary PASS vent 1\ation penetration seat with a permanent seat. ,

lhe only credible acciden* associated with this modification to the fatture of the charcoat adsorber, which would lead to a smatt release of radioactivity.

UrSAR Section 15.11 evaluates a simitar u.cident, a failure of the gaseous radweste system, inc\vding gross failure of the charcoal adsorber with an associated release to the environment. The off gas charcoal adsorber contains approximately 20,000 lbe of charcoat (Ref erence UFSAR Section 11.3.'J.S.9) white the PASS aosorber contains approximately 13 \bs of charcoat. Based on the rotative stres of the two adsorbers and the accident evaluation from UFSAR Section 15.11, any release from the PASS charcoat adsorber witt result in an insignificant amount of activity released compared to the off gas adsorber.

The PASS charcoal adsorber was constructed and tested to the roovirements of ANSI NE09/610. Therefore, the gross fattues of the housing is considered on

I SAFETY lALUATIONS l

[N01NEEMING DE$10N PACKAGES i Page 10 i

, Safety Evetuation No. 89-0116 NEV 1 (continued):

entremely unlikely event. The conduit penetration is seated in accordance with Detroit Edison Specification 3071 198, meeting the requirements of the WBAR. i In a PAS $ accident scenario it le highty probable that operators would require  ;

respiratory protection due to generet arct, airborne radioactivity. Thus even i without the PASS ventitation system, the operabitity of the PASS penet is not diminished. This change ente 6de component life by removing internet heat via forced air flow. It also reduces operational suposures by drawing any airborne contamination created by sampting teake or drips away from the personnel r drawing the sample.

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SAFETY EVALUATIONS tt4 GIN ([ RING DLSIGN PACKAGES Page t1 CAI ETY EVALUATION SlMAARY Safety Evaluation No 89-0121 Uf fM Revin ton No, b Referonce Documentt [DP 2187 REV A Secilon(e) N/A Table (s) N/A Figure Change (X1 Yes ( 1 No Title of Cimnget Sprinkler System Pressure Indicator and Drain Instattation in the Residust Heat Removat (RHR) Complex. Cebte Spreading Room, and Dieset Driven Fire Pump Room SOAMRY This modification instatted pressure indirstors and drains on wet pipe sprinkler systems in the hHR Compton, cable spreading room, and d i e s e 's driven fire pump room in order to facilitate surve111ance testing and bring the subject sprinkler systems into conformance with National Fire Protection Association (NFPA) Code Standard t3. The gages are for testing purposes and the drains are used for performing maintenance on the systems.

This design change does not change the function of these sprinkler systems.

The addition of the gages and drains mests the design requirements of the existing systemt and is bounded by previous pipe break analysis.

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VFEVY EVALUATIONS ENGINEERING DESIGN PACKAGES page 12 SAFETY EVALUATION SLMAARY Safety Evaluation No 89-0140 RI'V I UFSAM Revision No. f.

Reference Document l'DP 10S31 Sectlon(o) N/A _

Table (s) N/A l'igure Change (X) Yee  ! ] No Titta of Change Non-Interruptible Air Supply (NIAS) Aftercooter Drain Modification: N!AS Aftercootee Saf ety Relief Valvo Orawing and Calculation Corrections S M Y:

This change modified the N!AS af tercooter drain system f or each division by (1) replacing the original carbon stoet system with upgraded Seismic Category I and CA Level I piping, valves, Y-strainer, and condensate trap made of stainless stoet and (2) replacing the condensate trap bypass piping with a manual blowdown piping arrangenent. In addition, drawings have been revised to reflect the actual drain system configuration, CA levet classification, and piping group designation changes. As part of this modification, vwrlous drawing and stress calculation revisions were made to correct discrepancies between the des 10n and fletd instattations of NIAS aftercooter saf ety retlef valves F207A and B.

The new drain system witt eliminate the previous corrosion induced trap clogging problems and attow the use of a permanently instatted alternate manual condensate bloedown flowpath when the condensats trap is out of service. ,

This modification does not change the design basis or operation of the N!AS aftercooter drain system. Feature of this drain system does not create an accident which goes beyond the previously evaluated lose .of an entire division. The drawing and calculation changes do not impact the NIAS system design basis or operation.

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!h tW ETY LVAltlATION Stk94\RY i

Estety Fvatuation Noi 69-0154 UILAR Revision No. 6 Reference Doctraents l'DP 10714 Section(s) N/A 3

Table (s) N/A 1

Figure Change (X) Yes ( 1 No Title of Ctange: Instmustion of Primary Containment Water Level Ins t r umen t a t i on SLAS4WY :

A h toiled Controt Room Design Review Team was established to fulfill the requirements of NUREO 0737 to identify and provide improvements in the control room that of f er a high probabi\ity of improving plant safety by strengthening the man / machine interface. EDP 10?14, in conjunction with previounty inston ed EDP 8483 corrected one of these discrepancies by adding instrumentation which monitor drywen and torus pressure. This new instrumentation snown primary containment water tevet monitoring up to elevation 650 ft. Primary containment water levet monitoring r.apability up to the maximum floodeble level of the containment is needed for Fermi 2 Emergency Operating Procedures.

The design of the sensing line is consistent with the requirements of General Design Criteria 54 and 56 and Regulatory Guide 1.11 for instrument sensing lines penstanting prirna ry reactor containment to ensure primary containment integrity arid limit the potentist offsite dose below 10CFR100 requirements.

Seistnic instaustion of conduit, cables, recorder, transmitters, and instrument tubing ensures that the integrity of surrounding components or systems will not be impacted. The design of the drywell pressure sensing line is consistent with the design of other existing instrument lines penetrating primary containment and fans within the envelope of an instrument line pipe break accident scenario as described in UF SAR Section 15.6.2. Additionally, the ins t aMa t ion of the drywe n pressure sensing line does not connect to the reactor coolant pressure boundary and does not impact any ccmponent or system rotated to the safe shutdown of the reactor. Ut111:stion of the containment penetration for drywe t t pressure sensing is allowed by NRC approved License Amendment $7.

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SAFE 7Y EVALUATIONS ENGINEERING DESIGN PACKAGES Page 14 SAf ETY EVALUATION StMAW Safety Evaluation No: 89-0160 UFSAR Revision No. 5 Reference Document FDP 9417 Section(s) 11.7 Tabte(s) N/A _ _ _ _

Figure Change IX) Yes ( ) No Titte of Change: Onsite Storage Facility Modifications StMAMtY:

The on-site storage facility (OSSF) was originatty designed for handling, shipping. and storage of dry active weste (DAW) and asphetted waste contained in 55.gatton drums. An overhead erano system was instatted to cover essentially att areas of the building for remote operations with these drums.

Two things occurred which slightly modified the initial design input assumptions. First, parts of the OtSF are used for other (but similar) purposes such as DAW sorting and storage of temporary shield blankets.

Secondly, inasmuch as the asphalt system is not used, onsite vendors have been processing and storing (for short-term) redweste in large (170 ft ) tiners.

This modification will modify some of the OSSF watts enabling the aforementioned activities to be accomplished in a more ef ticient, safe manner with less radiation exposure. This_is necessary because the watts are too high for the movement of the 170 ft liners by means of the overhead crane.

The OSSF is not a Seismic Category I designed building, and won alterations witt not affect its structurat integrity. The changes in the OSSF have no direct interfaces with other plant systems or with nearby equipment which could cause malfunction of equiement impertant to safety. The OSSF contains no equipment.important to safety. The only equipment involved or interfaced with these changes are the OSSF crane, the forklift truck, and any portable vendor processing equipment. These changes will enable att of this equipment to work more efficiently. thus recucing the risk of an accident. Of f site doses to the generat public from OS$F operations will not increase as a result of these changes, and any internal radiation. environmental changes have been futty discussed and approved by plant radiation protection personnet. This modification does not change or add any pa t hway s for the release of radioactivity to the environment or for radiation unposure to the public,

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SAFE 7Y EVALUAT10NS ENQ1NEERING DESIGN PACKAGES Page 15 SAFETY EVAll1ATION St> AWRY Safety Evaluation No 90-0028 UFSAR Revision No. 6 Ref erence Doctanent: EDP 8321 ,

Sectionte) N/A Table (s) N/A Figure Change (X) Yes ( ) No Tit's of Changes FW Heaters 2N. 20. and 2S Level Transmitter Hoplacement SLM4ARY:

This modification replaced the originnt differential pressure type levet transmitters on foodwater heaters 2N. 2C. and 2S with displacement type transmitters. This change watt enhance plant rallability by reducing spurious heater alarms and control actions.

This modification does not change the operation or function of the affected f eed*a ter beaters. The new transmitters have the same quality levet, output signat, and toop scaling f actors as the original transmitters. Att downstream instrument loop components, setpoints. and functions remain the same. The amatt bore piping. insulation. and transmitters added to the second floor of the turbine building do not adversely impact the plant fire protection program.

SAFETY EVALUATIONS ENQlNCERING DESIGN PACKAGES Page 16 l

CAF ETY EVALUATION SLMMRY UFSAR Revision No. 6 Safety Evaluation No 9f.0029 Reference Documents EDP 8322 Section(s) N/A Tabte( e ) N/A l'igure Change (X) Yes I 1 No Title of Changet FW Hesters 3N, 3C, and 3S Level Transmitter Rertacement SUM MRY:

This modificotton replaced the original differential pressure type levet transmitters on feedwater heaters 3N, 3C, and 3S with displacement type transmitters. This change w111 enhance plant reliability by reducing spurious heater storms and control actions.

This modification does not change the operation or the function of the affected feedwater heaters. The new transmitters have the same quotity level, output signal, and toop scating factors as the or' 1 transmittees. Att downstream instrument loop components, setpoints, or, ,nctions remain the same. The smelt bore piping, insulation, and transmitters added to the second floor of the turbine building do not adversely impact the plant fire protection propram.

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SAFE 7Y EVALUA710NS ENQlNEERING DES!CN PACMGES Page 17 MFETY EVALUATION (AMMMY tief ety Evaluation Not 90-0030 UfGAR Rev&olon No, b Re f erence Doctanent : l'DP 8323 Sectlon(e) N/A Tabte(s) N/A Figure Change tX1 Yes [ ] No Title of Change FW Heaters AN, 4C, and 43 Level Transmitter Reptocement SlM MRY:

This modification reptaced the celginal differential pressure type tevet transmitters on feed *ater beaters 4N. AC, and 49 with displacement type transmitters. This change witt enhance plant retlebility by reducing spurious heater alarms and control actions.

This modification does not change the operation or the function of the offooted feedwater heaters. The new t"ansmitters have the same quotity level, output signat, and toop seating factors as the originot transmitters. Att downstream instrument loop components, setpointe, and functions remain the same. The smalt bore piping, insulation, and transmitters added to the second ficor of the turbine building do not adversely impact the plant fire protection program.

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SMEW EVALVATIONS l ENGINEERING DESIGN PACKAGES l Page 18 SAF E1Y EVALt1ATION $1& MARY Safety Evaluation No 90-0032 UfSAR Revisioti No. 5 Reference Document EDP 1t222 Section(s) 5.5 A.1.96 Table (s) N/A __

Figure Change (X) Yes i 1 No Title of Change: Outboard Main Steam Isotation Valve (MGIV) Internals and Valve Cover Replacement fR& MARY:

The vatve stem components and valve covers for outboard MSIVs 02l03F0280. C.

and D have been replaced to improve plant evattability and timit personnet radiation esposure through reduction of local teak rate testing (LLRT) fattures and by reducing the chance of a MSIV stem tallure. The changes to eacn valve includes (1) replacing the valve stem and poppet and pilot poppet assemblies; (2) implementing a live toed packing assembly and antirotational devices; (3) machining the valve cover / bonnet to attow popoet backseating on the -

cover / bonnet instead of the valve stem and to provide adequate clear ance for a the ese of stud tensioning devices; and (4) removing the valve, piping, and piping supports associated with the valve stem teakoff piping assembly as the new live load packing assembly eliminatos MS!V stem teakoff.

The reptocament of the outboard MSIV valve cover and internals, as wett es the comovat af the stem teskoff piping was carried out to improve valve operability, rettability, and maintainability. This modification does not affect the design basis, function, or sequence of timing of the MSIVs. The new parts are fatricated to the originnt design requirements, but are modified to improve packing performance and reduce the potentist for valve fatture.

Removat of the stem tenhoff piping does not create a new equiperent failure modo as a packing teak would reteese steam / water in the same general area as before.

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SAFETV D/At.UA7!0NS ENGINEERING DESIGN PACKACES Page 19 SAFETY EVALLIATION Sl8AMHY Safety Evaluation No. 90-0037 UFSAR Revision No. 5 Reference Documents EDP 1D610 Dection(s) 6.2 Tabte(s) N/A Figure Ctange IX) Yes ( ) No Titto of Change: Containment Nitrogen inerting. Venting, and Purging System Isolation Valve Upgrade SumMRY:

This modification upgraded the containment inerting, venting. and purging system Isolation valves by upgrading the limit switches and position indication ,

balance of plant power supply removing the non-Q pneumatic control solenoid valvesi and re pt, acing the oat pneumatic controt solenoid valves with a type that has a higher maximum operating differential pressure rating. This modification etso deleted the e=1 sting non. qualified automatic containment pressure control system. This change brings the containment nitrogen inerting, venting, and purging system isolation valves in agreement with the description in the UFSAR and resolves the concerns of NRC Information Nottee 88-24 and NRC Notice of Violations 89011 01A and 89011-02A.

With the exception of the automatic containment pressure control system that was deleted, the new control configuration functions identicatty to the previously analyzed system. The automatic mods of the containment pressure control system was not used and tre UFSAR, Technicat Specifications, and NRC Saf ety Evaluation Report do not take credit for it in any accident analysis.

Removal of the non. safety controle eliminates a condition wherein a common mode failure within the non. safety circuitry could have caused any of the isolation valves to open inadvertently. The new ccmponents and circuitry meet the requirements for class 1E safety grade components. This modification does not impact the stroke times of the isolation valves.

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SAFE 7Y EVALUATIONQ ENGINEfRING DESIGN PACKAGES Page 10 SAF ETY EVAltiA110N SLAWY Safety Evaluation Not 90_-0039 UFSAR Revision No. 6 Heference Document: EDP 11331 Cectlon(e) N/A Table (s) N/A rigure Change (X) Yee ( ) No Title of Change Rosetor Head Vent Sotenoid Valve Supply Power and Associated Equipment Removat SLMMRY t This modification disconnected the powse supply and removed att associated controt conter controle for reactor head vent t id valvee 021F403 and B21F404 Both valves are not functional because t ant piping upstream was plugged per EDP 10792. (See SE 89-0196 in Fermi 2 Safety Evaluation Summary Report, 1909.) This modification satistles human factor critoria by renoving equipment which to no longer required.

This modification does not affact the design. function, or operation of the reactor vesset because it is functionally equivalent to the previous configuration created by EDP 10792, l

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SAFEW EVALUST!DNS ENGINEERING DESIGN PACKAGES Page 21 TAFETY EVAlt1AT10N SLMAARY Safety Evaluation No: 90-0040 UFSAR Revision No. 5 Reiorence Document: EDp 9395 section(e) N/A Table (s) N/A Figure Change (X1 Yes [ ] No Title of Changes Addition of Remote Manual Close Control Switch in the Controt Room for Drywell Equipment Drain Pump Discharge Vatve SLM MRY:

This modification provided a remote manual close control switch in the control room for drywelt equipment drain pump discharge valve 01154F018 to comply with Reg. Guide 1,62 and the b4SAR. This modification replaces an interim design change (PDC 9384) that located the remote manual close control switch in the relay room.

This modification does not change other existing functions of the valve. It enhances the operators' ability to quickly close the valve if automatic isolation faits. The. new components instatted by this modification are of the same quality as the existing cc.apotien t s . Single failure criteria for any postulated fatture of G1154F018 is stilt satisfied in that, for any postulated failure of the new switch, a redundant outboard containment isolation valve exists to carry out isolation for this drain path, l

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SArtTY EVALUATION 3 EN31N(ERINQ D[$104 PACKAGES Pope 22 MFETY EVALLMTION SLa# WRY Safety Evaluation Nor 00-0041 UFSAR Revision No. 5 Ref erence Doctonant s (Dp 767t Section(s) WA Table (s) 6.2-2 Figure Change (X1 Yes ( ) No lit ~1e of Chs.nget Scram Dischargs Volume (SDV) Drain Valve Replacement and Addition of Manual Block Valves and Test Taps DLhMa,ny g This 'nedification roptoces the SDV inboard drain valve C1100F011 and SDV outboord drain valve C1100F181 and insta n s manuet block valves, test taps, and test tap bicek valves to f acilitate locat look rate testing (LLRT) of the SDV vent ano drain isolation vehen. The originat SOV drain valves have experienced emeessive leakage in three consecutive as-found LLRTs. The resultAht refurblehment has resulted in outage delays. The new drain valves provido e tighter shutoff capability. The SDV vent and drain manuat block valves, test tape, and test tap block valves have been added to facilitate individual valve LLRio. This win anow testing with the valves isolated from the control tod drivs system (CRD) and wiM result in improved trouble shooting capability $nc moes efficient testing and valve rework. The manual block valves and test tap block valves are designated as tocked vah e s and t.r e controMed under locked va\ve noministrative controle.

The BDV power supply, controls, and indicators are not altered or impacted by this modification. C1100F011 and C1100F181 use the same pneumatic supply for openinD and spring closure as the original valvcs. This modification maintains and satisfies the design requirements of General Electric Design Specification 22A6249 Revielen 3 Data Sheet s 22A6249AD Revision 10 and the Detroit Edison Specifications. The of f ect on drain flow capacity due to the replacement of the drain valves end instaM ation of the manual block valves was evaluated in a design calculation. It was concluded that although this modification reduces the crain- flow capacity by several gations per minute, it does not adversely impact SDV draindown capab nity. The effects of leaving a test tap block valve or leaving a drain or vont manual block valve closed were evaluated Leaving a test tap block valve open would result in the leakage of reactor condensate. This leakage is no different than the leakage that could be experienced if other manual valves in the CRD system are left open and, therefore, the addition of the two test tap block val,ves does not significantly change the probability of any one vab e tseing left o>en. Leaving a drain or vent manual block valve closed will impede condensato drainage. However, redundant SDV instrument volume levet trantmitters will detect the increasing level and alert the operators or initiate a reactor scram if corrective action

SAFEVY CVALUATIONS CNGINEERING OCS16N PACKAOCS Page 23 Safety Evaluation N. 90-0041 (continued):

r is not taken to restore deelnepe. The LLRT test procedures, locked valve program, and routine operator rounds provide multiple administrative coverage to ensure that the $DV vent and drain block valves are open and the test top shutoff valves are closed and capped.

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SAFETY [ VALUATIONS ENQlN((RIFl DESIGN FACKAGES Page 24 SAFLTY EVALUAtlON Sl W HY e

Safety Lvaluation Noi 90 0044 ftiV 3 UFLAR Revision No. ti Reference Documents EDP 11044 Section(s) N/A labte(s) 8.3 1 figure Change ( l Yes IX) No Title of ChanDet InsteMatton of a Dedicated 20aY/120V Maintenance power Oletribution System in Conteinment SUh4WIY This modification provided a permanent 208Y/120V maintenance powee distribution system tr. containment . The system provides two separato distribution networks!

one within the drywell and the other in the secondary containment. The new system consists of (2) AB KVA transformers with their associated raceway / cables and distribution networks. The primary power sources for the drywen and the secondary containment are Southside/MCC 72E-3A and Nor t hs i de /1ACC 728-4A.

respectively. An alternate source of power is provided utstiring a transfer switch and prefabricated cables capable of tying into the existing 400V maintenance distribution system. The system is de-energired when not in use and dorated from 100 emperes to 70 amperes when used during startup and power operations.

This system has no functional purpose related to or intseface with any safety I systems. AM components are sized and protected so that their electrical ratings are not exceeded. A fault within this system wi n only affect l

maintenance activities and wiu not de-energire other electricot equipment.

The instaustion is non-0 with the exception of the containment penetration 7

which is OA levet t. This change is in compliance with Hog. Guide 1,63. The l use of double fuses and de-energisetton is in accordance with the UFSAR. Plant procedures administratively controt doenergitation and dorating of the system.

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ILAFETY EVALUATIONS ENQlNCERING DCSION PACKAGES Psgo 26 LAFClY LVALUATION DUPA4ARY Safety Evaluation Nos 90-0046 UFMR Revision No. 5 Reference Documents IDP 7703 Section(s) N/A Table (s) N/A Figure Change ~X) Yes [ ] No Title of Ctange Addition of a Body Feed System to the Condensate Polisher Domineratiser System btASAARY:

This modification adds a body feed system to the condensate polishing comineralizers. This consists of adding a mining tank filt line, mixing tank drain piping, and individual vesset influent feed tubing. Body fonding consists of metering smett amounts of datineentirer resin to the filter columns (septs) during domineralizoe operation. Body feeding compensates for precoat imperfections, promotes depth fitiration, and reduces the differentist pressure buildup rate. As a result, filter performance is improved and domineratirer precoats are reduced.

The addition of body feeding is within the design resin loading of the domineratirers because the initial precoat tonding is lower (0.15 pst vs.

0.20 psf ) . The loss of the body feed system will not impact any previously evaluated accident analyses or affect safety related equipment. If interlocks fait and resin is continuoustv fed to the domineralizers, the vessels will be  !

removed from service due to h y t differentist pressure. Any pipe breaks in the body feed system witt not impact the reactor butiding teskage analysis. The medition or loss of the body feed system does not change the technical specification chemistry limits.

SAFETY EVALUATIONS

[NCINEERING DESIGN PACKAGES Page 26 LAFETY EVALUATION SUBAMRY E,afety Evaluation No: 90-0054 REV t UFLAR Revision No. 5 Reference Document IDf' 11429 Sectton(e) 6.7 Tabte(s) N/A Figure Change IX1 Yes ( 1 No Title of Changet IAodification of the High 70 int Vent on Division 1 RHR Return piping 1AA0AARY:

This EDP modifies the high point vent by removing one of the two vent isolation valves. s hor t e.31ng the remaining pipe spools, and modifying the sock-o-let fittet weld to reduce the susceptibility to vibration tonding.

The proposed design meets the dvat contalhment barrser design code requirement through the use of one isolation valve and a threaded pipe cap. The iO CFR 50 Generat Design Criteria 55 and CS provisions for primary containment penetrations and reactor coolant pressure boundary design are met through the use of; (1) a manuet isolation valve and a threaded pipe cap on the LLRT vent connection, and (2) the RHR system inboard containment isolation valve and the vent isolation valve.

This modification has no effect on the venting operation. The alternating stress tevels will be reduced by a factor of approximately three. The classification of the piping beyond the remaining vent isolation valve is currently ANSI 031.1, 1500s rating and will remain so in the revised design.

An effect of removing the vont isolation valve is a reduction in the administrative controts required to ensure vont closure following use since

' ttere is one less berrier between the header piping and containment. The existing administrative controls on vatve lineup and can verification assure leak tightness.

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SAFE.tv EVALUATIONS EN0tNEERlNQ DEO!QN iACKAGES Pena 21 l

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l Sof ety Entuation No! 90-0061_FEV 1 UFSAR Revision No. 6 Ref ererwie Doomant s EDP 7008 Section(s) 9.5 Tabte(si N/A F10ure Change ! J Yes [X] No l

Title of Changes Replacement of the Sound Powered Headset communication System with a Telephone System

$^m Y:  :

The purpose of this modification 18 to convert the existing sound powered headset system to a telephone syttem! provide additional telephone jacks, telephones, and sound proof telephone boothat pesvide additional Hl.com public address handsets, amplifiers, and loud speakeret and improve t. , radio and Hi Com communicatione in the kontrot room. Tais modification provides espanded communications coverage and enables the r,lant operators and lac technicians to condwet their activities more efficiently.

This change does not e f f ec t - t he operation of safety related equipment or systems and does not 4mpact- omisting ecotdent analyses. 1his modification pertains to non-Q systems and_does not impact the fire protection or appendia R criteria. The direct communication system between the controt room and the -

ref ueling Mat f orm required by the technical specificatione is not effected by this modification.

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SAFETY EVALUQTIONS ENQ1N8ERING DESIGN DACKAGES Page 28 TAFE1Y EVALUATION FAMMJ1Y Safety Evaluation No: 90 0088 UFSAR Revision No. N/A Reforence Document , __ E.Dp 10376 Section(e) N/A Tabte(s) N/A figure Change ( 1 Yom [X) No Title of Change Refusting i tat f or power / Control and C NLnuni c a t ione Cable Replacement SLAS M RY:

Yhis modification replaced the power / control cable and the communications cable on the ref usting plat f orm with a single composite cable that bandtes both the pletform power /controt and communicatione functions.

This modification does not create en unreviewed saf ety question. Thees in no potential for fust damage or radioactive release due to the entent of patential combustibility of cables. Fire hazards are not created because the location is already considered a tight combustible toading area and the cable la esparated from other cables. No fust damage or accidental critl%11ty will occur.

During refueling, the fuel platform will fall "as in" or, s toes of power.

Dueing normat omration, the refueting platform to de-energized and reactor vesset head is instatted.

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SAFETV EVALUATIONS l EN0!NEERING DES!CN PACKAGES l Page 29 l MFETY EVAlt!ATION SLMMRY Safety Evaluation No 90-0097 UFLAR Revision No. 5 Ref erence Docurnent : EDP 10452 _

Sectlon(e) G.2 l Table (s) 0.2-2) G.2 133 0.2 16 Figure Change (X1 Yes ( 1 No Title of Change: Main Steam lootet ton Velve Leakage Control System (MSIVLCS)

Isolation Valve Replacement SLA N M Y:

This modification reptaced the original sotonoid operated MSIVLCS isolation valves 021F433, F434, F437, and F438 with solenoid piloted bottons seated air operated valves. To accomptish this valve changeout four solenoid operated air ptivt vatves two non-interruptible lhotrument air system (NIAS) isotation vatves, PE000F1007 and P6000F10001 ir.d the associated piping, tubing. conduit, cable, c3M associat ed support s were instetted and four snubbers were deleted.

The 691VLCS isolation valves weec replaced because all four had previously f atted the required functional testing and F434 had fatted a local teaA rate test. Tiais resulted in increased maintenance .osts, entended plant down time, and incesseed personnet emposure.

This modification enhances system operability a r.d reliability. The modification has been designed in accordance with D$troit Edison approved i perscedures and specifications, There is no change to the system design basis, f usstion, or sequence of operation. The effect of a single fatture which resotts in the failure of one isolation vatve to open or close does not reduce MSIVLCS redundancy or re11 ability, A mal f unc t ion o f the components instatted by this destga does not impact the N!AS system, the aesociated power suppline, control circuits, or any adjacent woutpmerv important to safety, I

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BAFE7Y EVALUAVIONS ENGINEERING DES!QN PACKAGES Page 30 SAf E1Y LVALUATION SLA4MRY Safety Evaluation No 90-0099 UFSAR Revision No. 5 Referonce Document ,EDP 11633 Section(t) 9.4 ,

Tabte(s) N/A Figure Change ( l Yes IX) No Title of Change: Recirculation Pump Motor Generator (MG) Set Cooling Unit High Temperatura Trip Setpoint Change St& NARY:

This modification changes the trip setpoint of the recirculation purnp MG set cooling units from 106 F to 126 r in order to prevent unnecessary trips of the cooling units. Normat cooter outlet temperatures are approximately 100 F.

This modification has no effect on other equipment or MG set instrumentation and topic. AC set motoe temperature wilt stin be maintained wett below its high temperature alarm and trip setpoints. By attowing the cooling unit s to operate tonger, their ability to remove weste heat is improved and MG set performance is enhanced. ,

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SAFETY EVALUATIONS ENGINEERING DESIGN PACAA0ES Page 31 SAFETY LVALUATION SLA&RRY Safety Evaluation Not _90-0105 UFSAR Revision No. 6 Reference Documentt (DP 11260 Section(s) N/A Tabte(s) N/A figure Change (X1 Yes ( 1 No Title of Change Reactor Water Cleanup (RWCU) Filter Deinineratirer Efftvent Flow Controtter Replacement SUw wtY:

This modification replaces the pneumatic recording control stations C33R174A &

D with new pneumatic controllers. Replacement of the recording pen ink in the old controtters caused inadvertent changes in fitter dominera11aer flowrote which resulted in unnecessary dumping of dominerettger resins and nute&ncs alarms in the control room. The new contratters have independent auto and manual centrot units with futty balanceless and bumpless transfer, a highty visible seats display, and no recording function. Operations and Chemistry personnet requested elimination of the recording function because it does not paovide vitat information.

This modification does not change the function of the controtter or system and does not adversely affect any component or system related to the safe shutdown of the reactor. Removat of the recorder function enhances system performance because the recorder maintenance activities which caused inadvertent changes in domineralizer flowrote have been eliminated.

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L SAFE 7Y EVAi,UATIONS ENGINEERING DES!GN PACKAGES Page 32  ;

SAFETY EVALUATION SUWMRY Safety Evaluatton Not 90-0109 REV 1 UFSAR Revision No. 5 Reforoneo Document: EDP 11300 Secilon(e) N/A Tab \e(s) N/A Figure Change !X) Yes ( l No Title of Cfunge: Installation of a Second Rehester Seat Tank (RST)

SLMAARY This modification instaued an additional RST and the required piping, components, and essocistsd controls to provide a separate PST for each moisture separator reheater (MSR). The west MSR now dretn6 to the original (north) RST and the East MSR drains to the new (south) RST. This modification also modifies the north RST to accommodate the new piping configuration. The originat conterline drain notates are carped and a new noaste has been instatted at the top of the tank. A noarie and vent line to the SS feedwater heater outraction 11ne was added. Saffles were instatted inside the tank. The north RST differential crossure type levet t rane at t lers which control the normat and energency drains were replaced with displacement type transmitters.

The RST \evet starms and indication test pushbutton were etso eliminated.

This modification eliminates RST tevet instability and/or tank blow through at high power levels and during turbine intercept valve testing. The addition of a second RST doubles the capacity of the drain system. Relocation of the drain notate ensures that a highee drain level can be maintained without covering the drain nozzle. The replacement of the tevel transmitters makes the \evet control system less sensitive to wave action and turbulence in the RSTs resulting in less severe levet transmitter output fluctuations and fewer erroneous signals. The installation of the baffte plate reduces wave action.

This modification does nat modify the MSRs or MSR drain flow rate. The piping

-is designed in accordance with applicable codes and Detroit Edison specifications. There is no impact-on the previous pipe break analysis in the UFSAR. The addition of the second RST does not-effect the passive steam bypass system. The MSR drain lines are stred to carry the same drain flow as the previous configuention. An analysis of reheater bypass flow s hows that, for the 0%. 13%, and 26% flow bypass conditions consicered, the reheater f\ows lead to lov,er peak fust pin heat fluxes than those obtained by using the reheater flow pattern assumed in the bypass flow analysis in the UFSAR. Analysis of the effects of routing the RST vents to the $N and SS feedwater heater outraction steam tines indicate that the extraction flow rin increase by 3% and the effect is, therefore, considered negligible.

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SAFETY EVALUQTIONS ENQ14EERING OESIGN PACKAGES Page 33 SArETY EVALUATION CA8AMRY Safety kvaluation No: 90-Ott4_REV 1 UF FAR Reviolon No. 6 _

Referente Documents EDP 11190 Sectionto) 9.1 Tablete) N/A Figt4ce Change IX) Yes i 1 No Title of Change New Fuel Uprighting Stand Pelocation f'Jhs4ARY:

This modification retocates the fust uprighting stand 1.25 feet north and e f eet west in order to bring the stand within reach of the new fust transfer crane, This will attow the new fust crane to be used to transfer fuel bundles from the new f uel uprighting stand to the new fust inspection stand. Six new r

floor anchors are instatted, a UFSAR Figure change has been made to delete the abandoned or non-existant anchor tocations, and two UFSAR sections have been

. revised to rename the "untoading stand" as the "new fuel uprighting stand".

This modification does not atter the safety function of any other plant system

or equipment. The modification does not impact the heavy toad analysis for the 6th floor RB starage areas, structures or overhead crane es the new fust uprighting stand weighs less than 2000 lbs and is not considered a heavy load

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EAFETY EVALUA710NS ENGINEERING DE81GN PACKAGEb Page 34 SME1Y EVALUATION Sl84MRY Sasety Evaluation No 90-0115 REV 1 OfLAR Revision No. N/A l

Pef erence Doctanent EDP 1_I__282 Section(s) N/A Table (s) N/A ,

Figure change ( ! Yes (X1 No Titto of Change Reheater Seat Tank lintrot Center COP H11P805 Changes and Additions DLaAWRY:

This modification made the f ollowing chunges to the controt center operating penet COPH11P905:

1. Provided the controts and indication f or a second Reheater Seat Tank being added by EDP 11300.

2. Addressed a portion of HED-455 by removing a number of unused controle (push-button / indicators) for valvea removed or abandoned in place by PDC 9357. PDC 9359. EDP 8938. EDP 10778, and EDP 11300.

This modtfication does not change the function or operation of the f oodwatee heater drain system or its associated components. This system does not affect the operation of any systems required for the saf e shutdown of the plant nor does it change any scetdent scenarios credible to this controt room panet.

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SAFCTY EVALUA?!ONS

[NQ1NE(R!NG DESIGN PACKAGES .

Page 36 SAFETY EVALUATION Cl84MRY Safety Cyatuation No 90-0116 ,

UFSAR Revision No. 5 Reference Documents ._EDP 11t>02 Section(s) 6.2 Tabte(s) N/A ,

Figure Change (X) Yes 1 1 No Title of changei Division !! Residuet Heat Removat Return Piping High Point Vent Modification S M Y:

This modification removes the second isolation valve end replaces it with a threaded pico cap. The purpose of the change is to reduce the vibration induced stress tevets of the vent tino end reduce the probability of fatture at the wetdment of the vont line and 24" RHR pipe " sock-o-let". This design does not s,atisfy the requirements of UFSAR section 6.2.4.4.3 which states that test, vent, and drain connections on the Class 1 system (which are part of the containment boundary) are provided with at toast two iso",ation vetves and are seated with a threaded pipe cap. However. 10 CFR LO Generat Design Criteria 56 and 58 are met because penetration to primary containment isolation in accomplished through the use of the commining vent valve and pipe cap; reactor coolant pressure boundary isolation in accomplished through the use of the remaining vent valve and the RHR system inboard containment isolatten valve.

This modification does not change the saf ety f unction or operation of the RHR system. Any fattures introduced by this modification are t>ounded by the aman break anatyeis. Administrative controts exist to ensure validation of vatva closure after use.

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SAFE 7Y EVALUATICNS ENGINEERING DESIGN FACKAGES Page 36 SAFETY EVALUATION SUPARRY Safety Evaluation No: 90-0115 UFSAR Revision No. N/A Reference Document: EDP 11819 Section'*) N/A Tabte(s) N/A Figure Change I 1 Yes (X) No Title of Changes High Pressure Coolant Injection (HPCI) High Steam Flow Transmitter Time Characteristics Modification

$l3AMRY:

The purpose of this modification is to increase the reliability of the HPUI high steam flow isolation instrumentation loop by filtering out undesirable procesa noise and oliminating noise as a cause of spurious HPCI isolatlans. A capacitor has been added between the output of the transmitter and the input of the associeted trip unit. The new capacitor adds 2 seconds to the loop response time but overalt loop response time' is well b. low the Technical Specification response time requirement of 13 seconds.

This modification does not change the function of the HPCI system. The analyzed basis for a high energy line break in the HPCI steam supply line is not compromised by this change because no credit is taken for the operation of the high steam flow togic to mitigate a steam supply line break. Failure of the added components does not affect any accident or transient analysis and can be detected during shift surve111ances.

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SAFETY WAL"ATIONS ENGINEERINO DESIGN PACKAGES Page 37 SAFETY EVALUATION St#AW(Y Safety Evaluation No 90-0121 REV 3 UFSAR Revision No. 5 Referonce Doctamenti EDP 11274 Section(e) N/A Tabte(s) N/A Figure Change 1X] Yes ( 1 No Title of Change: IristaMation of Station Air Hender Pressure Instrumentation for the Control Room Sth5Aa,ny:

The purpose of this modification is to provide the control room operators with a more distinct indication of station air preemure by installing station air header pressure indication in the control room. The modification consists of the installation of a pressure transmitter, u pressure indicator, and the essociated tubing and cabling. The original air pressure indication, consisting of Division I an( Division II NIAS compressor discharge pressure.

was reviewed by the Detailed Control Room Design Review Team and found to be an inadequate indication of station air pressure, This concern is documented in Human Engineering Deficiency HED 1177.

This modificat10n does not adversely affect any component or system related to the tafe shutdown of the reactor. The indication provided by this modification is not required for any safety functions nor is.it requiaed to operate after a design. basis accident. The componenta used for- this modificat. ion are similar to other components in the plant that have proven reliability.

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SAFETV EVALUATIONS ENGINEERING DE01CN PACKACES Page 38 SAFETY EVALUATION SLMMRY Safety Evaluation No 90-0173 REV 1 UFSAR Revision No. 5 Reference Document: EDP 11266 Section(s) N/A Table (a) N/A Figure Change (X) Yes ( ) No Title of change: Control Room Multipoint Recorder Reptacement StA4RRf:

This modification resolves part of Lead Human Engineering Discrepancy HED 907 by replacing the 18 original L&N Speedomax multipoint recorders with 13 Westronics Series 3000 and DDR10 recorders. The new recorders provide a legible trace and have an easily readable display, point sel,ectability, and. tow maintenance. Due t o - t he physical size of the replacement recorders, some recorders that monitor similar variables have been combined into one recorder to provide room for installation and free up control room operating panet space for future instrumentation.

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This modification will not adversely affect the function of existing equipment or their perfoemance as originauy designed. Performance is enhanced by the improved readability, added point selectability, and reduced maintenance,

SAFETY EVALUATIONS

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ENGINEERING DESIGN PACVACES Page 39 SAFETY EVALUATION SuuAnay Safety Evaluation No: 90-0125 UFSAR Revision No. _

5 Reference Document: EDP 11803 Section(s) 10.2 Tabte(s) N/A i

Figure Change (X1 Yes [ ] No j Title of Changet Low Point Drain Installation for Feedwater Heater 4 Extraction Steam Line SLM4ARY: )

This modification instatted a drain on the extraction line of the 4N feedwater j heater between check valve N3000F402A and the heater at a previously I undrainable low point. A walkdown of the feedwater heater room during plant ,

outage 90-04 revealed damage to the 4N heater support pedestal and extraction steam line pipe support N30-3199-017. It was also noted that the 4N feedwater l heater shifted north approximately 2.5 inches. The cause was determined to be a fluid transient in the extraction steam line.

The new drain line is connected to an existing drein between N3000F402A and

- N3016F603. Both drain lines are isolated by a common isolation valve. To avoid having an open bypass around check valve N3000F402A, a check valve has been instaMed in the nem drain valvo.

This modification has no effect on any o f - t he UFSAR - chapter 15_ accident _

analyses. However, since there are no provisions for testing the new drain line check valve as required in UFSAR section 10.2.2,6 the impact of water induction due to a failure of the subject check valve was anatyred. It was determined that the open area of the subject check valve disc would not attow a significant amount of drains from the 4N feedwater heater to back flow into the low pressure turbine and that no turbine overspeed hazard exists.

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SAFEVY EVALUA110NS ENGINEERING DESIGN PACKAGEU Page 40 SAFETY EVALUATION SanMRY Safety Evaluation No: 90-0136 UFSAR Havision No, 5 Reference Document: EDP 11527 Section(s) N/A Table (s) N/A Figure Change [X) Yes [ ] No Title of Changes Main Steam Line Pressure Tap Removat SLBJMARY:

The purpose of this modification is to reduce the potentiel for main steam tino tenkage by removing the pressure tap source valves and copping 4 nuclea-boiler system pressure taps and isolation valves on each of the main stof-Lines. These taps were originally used to obtain pressure data during startup testing and are no longer used. This modification effects the 3/4 inch instrumentation tines located between the outboard main steam isolation valves (MSIV) and the third MSIVs downstroem of the main steam drain lines.

No new equipment is added by this modification. This modification will not affect the operation of the MSIV teskage control system or the main steam lines drains. In.the event of a LOCA, att potential MSIV teskage originating from primary containment will be contained. The potential for steam line leakage is reduced because removal of the pressure tape reduces vibratory stress loads at the tap connections. This destgn satisfies UFSAR stress limits for pressure, weight. seismic, and transient tondings.

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SAFETY EVALUATIDNS ENGINEERING DESIGN PACKAGES Page 41 SAFETY EVALUATION SUIMARY Safety Evaluation No: 90 0137 UFTAR Revision No. 5 Reference Documentt EDP 8972 Section(s) N/A Table (s) N/A Figure Chango IXl Yes ( ) No Title of Chr.nge: Post Accident Sampling System (PASS) Valve Cable Rerouting and Reterminating SlMAARY:

This modification rerouted and roterminated four cables that are part of the control scheme for the PASS residual heat removal liquid sample valves P34F402A and P34F402B to ensure that each valve is controlled and powered from its respective divisional control panel. Originally, the valves were wired such that the Division I vatve was controlled and powered from Division 11 panel H11P618. The Division 11 valve was controlled a r.d powered from Division I panel H11P617 This modification ensures that the sample valves are controlled from the same division as the division being sampled. This modification also retage the valves and associated limit switches to agree with other PASS system components ano updates UFSAR figure 11.4-8 to show the corrected PIS numbers.

This modification does not change the function of the PASS system as described in the UFSAR. Plant operation is enhanced in that the samole valve controls for each division RHR sample train are on the proper control panel.

SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGE 5 Page 4?

SAFETY EVALUATION e M RY Safety Evaluation No: 90-0138 UFSAR Revision No. 5 Reference Document: EDP 9922 Section(s) 7.2 Table (s) N/A Figure Change ( ) Yes IX) No Title of Change: Reactor Protection System (RPS) Electric Protection Assembly (EPA) Logic Card Replacement a StANARY :

The purpose of.this neodification is to enhance the performance of the EPAs by replacing the originat design EPA logic card with an improved logic card and modification kit. This modification addrescos the concerns of G.E. SIL #496, revision 1 and RICSIL 8026. EPA performance and reliability will be enhanced by the followingt

- Elimination of spurious trips from causes internal to the logio card

. Resolution of the IC chip lockup concern described in RIC51L #026

- Provision for a connector interface to the logic card

- Improved access to test points needed during routine catibration of the Logic card

- Reduction of the number of mechanical cycles to which the circuit breaker is subjected by adding a switch to disconnect the circuit breaker undervoltage release (UVR) coit during logic card calibration

. Rated voltage is provided to the UVR coil.

As a result of this modification, the time delay continuous adjustment range

, has changed from 0.2 - 3.6 seconds to 0.3 - 3.6 seconds.

This modification does not change the function of the EPAs or RPS. It enhances RPS availability and improves EPA performance. The . change in time delay continuous adjustment range is still within the anatyred limits and is, therefore, acceptable.

SAFETY EVALUATIONS ENGINEERING DE31GN PACKAGES Page 43 SAFETY EVALUATION SLMAaJgy Safety Evaluation No: 90-0139 UFSAR Revision No. 5 Reference Document: EDP 11068 Section(s) 7.1; 7.23 7.6 Table (s) N/A Figure Change ( ) Yes (X) No Title of Change: Process Computer Core Monitoring Software Rep 1,acement SLMMRY:

This modification replaced the process computer system core monitoring GEXL+NSSS software with Generet Etoctric 3D-MONICORE software. DEC work station computers were etso added to support the 3D-MONICORE system.

This modification does not change the process monitoring system. The DEC work station computers do not provide information to other plant systems. The absolute accuracy of the 30-MONICORE core physics model has been established by comparing its calculated results with gamma scan measurements carried out at Edwin I. Hatch Unit 1 following cycles 1 and 3. The 3D-MONICORE uncertainties are covered by the safety limit margins employed in the process computer. In the event that the process computer is not available, the software operates from en onsite DEC Microvax C3800 with futt backup from a DEC Vax Station 3100. This software backup capability is a significant improvement over the former system.

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ENGINEERING DESICN PACKAGES Page 44 SAFETY EVALUATION e M RY Safety Evaluation No: 90-0142 _

UFSAR Revision No. 5 Reference Document: EDP 11889 Section(s) 6.4; 7.1; 7.3 Table (s) 9.4-2 Figure Change I 1 Yen IX) No Title of Change: Control Center HVAC (CCHVAC) Control Logio SLM MRY:

The purpose of this modification is to change the CCHVAC control logic to: (1) prioritire the recirculation mode of operation over the chlorine mode and (2) require the operation of_the mode select reset pushbut ton f or mode actuation.

The recirculation mode was prioritized to ensure that. in the event of a chtv.ine detectoe failure followed by a subsequent LOCA/ radiation release, the CCHVAC will automatically transfer to the recirculation mode. The former 1,ogic required the operators to manually initiate the recirculation mode. The mode selector logic was changed to require mode select reset button operation for au modes of actuation. This change provides a uniform method for mode

- initiation. The former logic did not always require the operator to press the reset button to initiate a mode. For some modes, the mode would be initiated when the mode.was selected. For other modes, the modo selector reset button was pressed to initiate the selected mode.

This modification does not add any equipment or affact the equipment within the system. The original intent of the CCHVAC mcde selection process has been maintained. The automatic initiation signals for both the chlorine and recirculation modes override any manually selected mode. The UFSAR accident analyses address the operation of the recirculation and chlorine modes individuany. A LOCA is not addressed when operating in the chlorine modo and a chlorine accident is not addressed when operating in the recirculation mode.

A single failure of a chlorine detector will not prohibit the CCHVAC recirculation mode from automatically initiating. Calculations show, that with CCHVAC in the recirculation mode,- all chlorine accident scenarios result in centrol center chlorine concentrations less than the toxicity limit of 15 ppm.

Centrol center personnel will be limited to 5 rems whole body or its equivalent consistent with the requirements of Generat Design Criterion 19 of 10 CFR 50 Appendix A.

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SAFETY EVALUATIONS ENQ1NEERING DESIGN PACKAGES Page 45 SAFETY EVALUATION SLMAARY Safety Evaluation No: 90-0147 UFSAR Revision No. 5 Reforence Document: EDP 9310 Section(s) N/A Table (s) N/A Figure Change [X] Yes [ ] No Title of Changer Removat cf Limit Switches on Reactor Water Cleanup (RWCU)

System Check Valves SLW ARY:

This modification removed the actuator and disc position limit switches on RWCU check valve G330F121 and the actuator position limit switches on RWCU system check valve G330F120. In addition, the controls and position indication for G3300F121 and the actuator position indication f or G3300F120 that are located in the controt ' room have been removed. The solenoid valve for G3300F121 has been removed and fittings and labels have been provided at the air manifold shutoff valve and check valve actuator tubing to identify the connection points f or the instattation of a pneumatic jumpar for future inservice inspection testing (IST). The position timit switches removed by this modification have

- been high maintenance items and their repair has raised ALARA concerns.

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This modificat1on does'- not change. the function of_ these check val,ve s . The controls, actuator position indication, and disc position indication 'are not 4 required for G3300F121. G3300F121 only needs an actuator when-it is required-to be open es part of a valve lineup for local leek rate testing. The actuator-position indication is not required f or G3300F120. G3300F120 only requires disc position indication for IST- program testing. Removat of the limit

- switches does not impact the seismic calculations for t he se valves as the weight of-each limit switch is less than 10 ounces. This is less than 1% of the total valve weight and, as such, is insignificant.

SAFETV EVALUAT30NS ENGINEERINO DESIGN PACKA0ES Pege 46 SAFETY EVALUATION SLM.1A3Y Safaty Evaluation No 90-0155 REV 1 UFSAR Revision No. 5 Refersnee Document: EDP 11591 Caction(s) N/A Table (s) N/A Figure Change (X) Yee ( ) No Title of Change Removat of Feedwater Check Ve've Limit Switches SLMAARY:

This modification made the following changes to feedwater check valves B2100F076A, B2100F0768. B2100F010A, and 82100F0108:

B2100F076A and B

1. The actuator limit switches were removed.

2. The actuator position indication was removed from control room operating panel C0P insert H11P603A504.

3. A connection was installed in the actuator mir supply piping to allow

'the use of a pneumatic jumper to facilitate actuator testing.

82100F010A and B

1. The octuator and disc position limit switches were removed.

2. The actuator solenoid valves were replaced with shutoff valves.

3. Local wiring and flexible conduit were removed and the resultant conduit and electrical box openings were plugged.

4. The control -indication was removed from control operating panel Cop insert H11PS03A504.

This modification removed unneeded indication for 821D)F076A and B and unneeded indication and controls for B2100F010A and B. It also enhances the maintainability of the check valves in that their limit switches have been high maintenance items.

SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES Page 47 Safety Evaluation No. 90-016b REV 1-(continued):

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.i This modification does not affact the check valves' ability to close on reverse l l

flow and does not -degrade their performance as containment isolation valves.

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-The valves will respond to the accident conditions as assumed in the UFSAR. l The B2100F075A and B position indication required by the IST program remains in place. This design reflects consistency with NUREG 0700 " Guidelines for Control Room Design Reviews" for controt placement on controt room operating panets. .

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SAFETY EVALUAT3ONS ENGINEERING DESIGN PACKAGES Page 48 SAFETY EVALUATION SW6MRY Safety Evaluation No: 90-0156 UFSAR Revision No. N/A Reference Document: EDP 11354 Section(e) N/A Tabte(s) N/A Figure Change ( 1 Yes (X) No Title of Change: Instrument and Controt (I&C) Instrument Line Snubber Removal SL54MRY:

This evaluation justifies the removal of 100 anubbers from 25 I&C instrument Lines. 28 of the snubbers have been replaced with struts and the remaining 72 snubber locations do not require support. The original piping stress analysis conservatively assumed excess flow check valve testing would occur at a reactor temperature of 546 degrees F. As a result, the originel pipe stress analysis predicted large thermal movements. Snubbers were installed to accommodate thermal movement and seismic loads. However, excess flow check valve testing is performed when the piping temperature is less than 200 degrees F (operational condition 4). Roanalysis of the above instrument lines, using 200 degrees F as the operating temperature, allowed the above snubbers to be remeved without overstressing the instrument lines or the remaining supports.

.The analysis also indicated that 28 of the above snubbers'would be replaced by struts. Elimination of these snubbers witt: (1) decrease the costs associated with the periodic maintenance, inspection, and testing of_ snubbers; (2) decrease the radiation exposure to personnet; and (3) decrease the potential for extended outages due to snubber failures.

i This modification has no adverse impact on plant safety no- does it not impact the function or operation of any interfacing system. -The ability of the subject instrument lines to f unction has not been degraded by the removat of these snubbers. The qualitative vibration assessment in UFSAR section 3.9.1.1.3 has not been effected.

SAFETY EVALUAVIONS ENGINEERING DESIGN PACKAGES rage 49 SAFETY EVALUATION SUE MRY Safety Evaluation No: 90-0160 UFSAR Revision No. 5 Referonce Document: EDP 11663 Section(s) N/A Tabte(s) N/A Figure Change IX) Yes ( ) No Tatte of Change: InstaMation of Reactor Bull. ding HVAC (RSHVAC)

Ductwork Catch Pan and Drain StAMARY t This modification added a duct section in the RSHVAC ductwork at the scrsm discharge volume LSDV) vent to RBHVAC duct connection. The ductwork section is made of stainless steel and equipped with.a catch pen and drain line. The drain line is routed from the ductwork to reactor building floor drain sume G11010076 via a basement levet instrument blowdown drain. An access panel, is provided on.the exposed side of the duct for decontamination purposes. P1,an t operators had observed water on the floor of the residual heat removat Division 1 heat exchanger room after reactor scrams. Analysis of the water indicated that the source was reactor condensate. Further invest 4gation revealed that the water was entering the RBHVAC duct work from the SOV. vent line. Inspection of the galvanized steet ductwork revealed significant corrosion damage.

Installation of - the stainless steel ductwork will prevent further corrosion

6. mage and the. catch pan and drain wil.1, prevent condensate spills.

i This portion of the R8HVAC ductwork is not required for safe snutdown of the reactor or for maintaining secondary containment. This modification does not change the operation or function of the SDV system, the RBHVAC system, or floor drain sump G11010076. The instatted components are structuratty mounted to prevent interference with safety related components or systems under design accident anf seismic conditions. The design-of tb- irgin line minimizes crud trapping and br* spot formation and provisions a r e ... ace for *1cshing the line L

if contamination is detected.

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SAFETY EVALUATIDNS ENGINEERING DESIGN PACKAGES Page 50 SAFETY EVALUATION SLANARY Safety Evaluation No: 90-0163 UFSAR Revision No. 4 Reference Document: EDP 11974 Seetlon(s) N/A Table (e) N/A Figure Change [X) Yes I] No Title of Change: Installation of Primary Containment Atmospheric Grab Sample Taps SLANARY:

This _ modification installs grab sample taps at the primary containment radiation monitoring skid H21P284 for the purpose of taking primary containment atmosphere grab samples. The location of the taps ensures that the manual grab sample rig will be automatically isolated from containment in the event of a LOCA. Each sample tap is equipped with a manual isolation valve and a plug.

The operation of the isolation valve is administrative 1y controlled by independent verification by its users. Since the sample taps are located in Division I, the Division II sample taps may bo used with strict administrative controls when the Division I taps are not available. The use of the Division II taps is justified in another safety evaluation.

Tha new design maintains primary containment integrity during any design basis accident es watt as providing grab sam?le capability during normal plant operations. This modification does net impact the reactor coolant boundary nor does it change the operation or function of the primary containment monitoring system (PCMS). The administrative controle placed on the operation of the sample valves ensure that any potential for breaching containment or rendering the PCMS inoperable due to the operation of these valves is eliminated.

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SAFETV EVALUATIONS ENGINEERING DESIGN PACKAGES Page 51 FAFETY EVALUATION SUWWRY Safety Evaluation No 91-0004 UFSAR Revision No. 5 Reference Document: EDP 70J2 Sectlon(e) N/A Table (s) N/A figure Change 1X) Yes [ ] No Title of Change: Turbine Building Closed Cooling Water (TDCCW)

Corrosion Control SMAARY:

This evaluation justifies the use of sodium mo11 bate / sodium nitrate as a corrosion inhibitor in the TDCCW system. An evaluation of the pure water application method (no corrosion inhibitor chemistry) used in the TDCCW system showed that only 5% of the system volume was flowing through the filter domineralizers and that turbidity levels were unacceptable. The evaluation concluded that chemical treatment of the TDCCW syst>m was required. To monitor the performance of the corrosion inhibitor, coupons were installed in the return flow path upstream of the host exchangers. The coupons are made of metals found in the TDCCW system.- Pipe tape were instatted in the supply and return headers to allow the use of a temporary portable filter domineralizer when the corrosion inhibitor cannot be injected into the system.

This modification enhances the reliability of the TDCCW system by constantly monitoring for system boundary degradation. The TDCCW system is not safety related and is not required for the safe s hu t down o f the reactor. If a pipe tap breaks and causes a breach of the TDCCW system boundary, the probloc vould be identified by the observation of opening makeup tank fill valve, low makeup tank level, and low system pressure alarms. The initiation of any of these alarms will alert the operator to take appropriate action to mitigate the loss

( of inventory. If a coupon should break free, the pump suction strainers will prevent it from entering the pump.

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SAFETV EVALUAVIONS ENGINEERINO DESIGN PACKAGES Page 62 SAFETY EVALUATION StMMRY Safety Evaluation No: 91-0006 UFSAR Revision No. 6 Reference Document: EDP 11577 Section(s) N/A Table (e) N/A Figure Change 1X) Yes i 1 No Title of Change: Residust Heat Removat (RHR) Air Actuator Replacement and Limit Switch Removal SLM MRY:

This modification replaced the air actuators and removed the actuator timit switches on RHR check valves E1100F050A and E1100F050B. The new actuators increase the check valve disc stroke from 22% to 68% open. This ensures that the check velve disc open limit switch will actuate when the valve is opened and provide open indication in the control center. The former actuator's stroke was too short to actuate the open timit switch. As a result, the position indicators in the control center showed no open or closed indication j when the check valve was opened. The removal, of the actuator limits switches eliminates unneeded indication from the control center.

This modification does not change the function of the check valves. The actuators are used for check valve testing only. The change in actuator nitrogen consumption does not impact the operation of the nitrogen system. The change in actuator weight does not impact the existing seismic analysis. Check vrtve position can still be observed in the controt center by using the disc position indication.

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ENGINEERING DES!CN PACKAGES Page 53 SAFETY EVALUATION St# MARY Safety Evaluation Not H @l0 UFSAR Revision No. 5 Reference Document EDP 11948 Section(s) 8.2 Table (e) N/A-Figure Change [ ] Yes IX] No Title of Change: 4160 V. Bus Voltage Low Alarm Voltmeter and Controller Rogsacement SulWARY:

This mool:4 cation replaced the contact making vottmeters and controllers for the Division I and II 4160 V. bus low voltage alarms with digital voltmeters and controtters. The internal components and the indicator for the original instrumentation is obsolete.

The new instrumentation performs the same function as the original instrumentation in that a degraded grid voltage condition alarms to attow the operator to take action bef ore an undervoltage bus trip actuates. The alarm setpoints and time delay rettings have not been changed.

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SAFEVY EVALUATIONS i

ENGINEER!NQ DESIGN PACKAGES Page 54 SAFETY EVALUATION SLA9AARY Safety Evaluation No: 91-0013 UFSAR Reviston No. 6 Reference Document: EDP 10442 Section(s) N/A Tebte(s) N/A Figure Change (X) Yea ( ) No Titte of Change: Peplacement of the SN and SS Feedwater Hester Condensate Overpesssure Thermat Retief Vatves SLANARY:

This modification replaced the UN and 69 feedwater heater condensste overpressure thermat cetief valves. N2000F302A and B. The original retief valves had a 40% blowdown range, This range wee considered unecceptable cecause it caused the relief valve setpoint to drift resulting in valve lifts during plant transients. The new relief - valves have e bloudown range of 20L The new vetven are equipped with a built-in travet stop to prevent vatvo over-travel and possible setpoint drift.

The setpoint, valve size, and mase flow rate of the new relief valve.s are the sama as the f orrner relief valves. There is no change to the design function or

-intanced system response to plant operuttrsg tranateato. . There are no failure modes or mech;nisms possible that'are obt assumed for the former volves.

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SAFETY EVALUATIONS ENGINEERING DESIGN pACXAGES Page 55 SAFETY EVALUATION St&4RMY Safety Evaluation Not 01-0024 UFSAR Revision No. 5 Reference Doaument EDF 12176 Section(s) 9A.4 Table (s) 9A.6.1-1 Figure Change (X] Yes 1 ] No Title of Changel Modification of Division I Switchgear Room Smoke Detection System St.MMRY:

This modification installs a single smoke detector in the ceiling of a Division II cable enclosure located in the division I switchgear room. The lack of a detector in this enclosure had been declared a limiting condition for operation and a roving fire watch patrol had been assigned as a compensatory measure.

Instellation of the smoke detector provides the fire detection instrumentation required by section 11.0 of NUREG-0798 Supplement 5. " Safety Evaluation Report Related to the Operation of Fermi-2" and eliminates the need for the fire watch.

The extension of the switchgear room fire detection system enhances safe pt. ant operation and shutdown. The installation of the new detector decreases the probability and consequences of.a fire as eval,uated in the UFSAR fire hazards analysf.s. The cable enclosure does not contain any equipment other then cable trays and conduit, Therefore, the only type of accident that.could occur in this room is a cable fault fire. The new detector and its associated conduits a^e supported in accordance with seismic II/I critaria.

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SAFETY EVAL.UAT}0NS ENQlNEERING DESIGN PACKAGFE Page 56 SAFETY EVALUATION SuleMMY Safety Evaluation Not 91 0025 __

UffJa Rwiston No. _ $

Reference Documentt EDp 12080 Section(s) 10.4 Table (s) N/A figure Change ( 1 Yes IX) No Title of Changet Main Condenser Retube StADMRY:

This evaluation justifies revising the description of the main condenser in the

-UFSAR due to the retubing modification perf ormed during the second refueling outage. The original admiralty brass and 70/30 copper / nickel tubes were replaced with titanium tubes. The original tubes experienced stress corrosion cracking, tube pitting, and were a significant contributor to the amount of dissolved copper in the f eedwater system. The interaction of copper with the zircatoy fuel rod cladding results in crud induced localized corrosion (CILC).

The corrosion can occur when the reactor is operated above 85% power.

The main condenser is not required for safe shetdown of the reactor or operation of emergency safety features equipment. The new tube material does not c hange the f unction of the condenser. Removal of the orir inal condenser tubes removes a major source of copper and is expected to reduce the possibility of CILC. Reduction of CILC mitigates the potential for fuel failure caused by corrosion of the (vet rod zircoloy cladding. The titanium l tubes do not react with the aircaloy cledding. The effects of condenser uplift.

tube vibration, and tube sheet stress were analyzed and the results were f our.d to be acceptable.

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SAFE 7Y EVALUAVIONS

-ENGINEERING DESIGN PACKAGES Page 67 SAFETY EVALUATION SLANARY Safety Evaluation No: 91-0027 UFSAR Revision No. 5 Reference Document: EDP 12 t 22 Section(s) N/A Table (s) N/A Figure Change (X) Yes ( ) No Title of Changes Main Steam Safety Relief Valve (MSSRV) Solenoid Valve Modifications

SUMMARY

This modification changed the solenoid arrangement for MSSRVs 82104F013A through d, J through N. P. and R from a double to a single solenoid valve arrangement. The second solenoid valve was non-f unctional and did not affect the operation of the MSSRVs. It was originally added to accommodate the proposed General f.ir,ctric Prompt Relief Trip System (PRT). The PRT system was never implemented et Fermi 2. Removal of the spare solenoid valves is a maintenance enhancement in that it eliminates unnecessary EQ refurbishment Oosts associated with each spare, non-functional solenoid valve.

The MSSRVs continue to function and operate as originally designed and as described in the UFSAR. This modification does not affect the qualification of the MSSRVs because the solenoid valves are qualified in both the single and dual solenoid configurations.

SAFETY EVALUATIONS ENGINEERING DESIGN P_acKAGES Page 58 SAFETY EVALUATION *NY Safety Evaluation No: 91-0033 UFSAR Revision No. 5 Reference Document: EDP 12175 Section(s) N/A Table (s) N/A Figure Change (X1 Yes [ ] Ho Title of Change: A'. ternate Drywell Head Storage Locttion on UFSAR Figure Sl544ARY:

This evaluation justifies revising UFSAR Figure 3.0-31, Sheet I to ittustrate a, alternate drywett head storage location over the dryer / separator storage ptet. The support assembtion, consisting of beams and plates located in four places at the edge of the storage pool, are shown in Detroit Edison drawing 6C721-2340, revision U.

Storing the drywell head at a location other than the designated 1.aydown area on the refueling floor does not involve any equipment or affect the performance of any system. The function of the dryer / separator and dryer / separator storags pool are unaffected by storing the drywell head over the pool. Seismic analysis shows that the support assemblies wilt not move and are capable of handling drywell head loads during a design basis seismic event. Compl,iance with NUREG.0612 ensures that a toad handling accident will not occur.

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SAFETY EVALUATIONS ENQ1NEERING DESIGN PACKAGES Page 59 SAFETY EVALUATION SLMMRY Safety Evaluation No: 91 0034 UFSAR Revision No. 5 Reference Document: EDp 11963 Section(s) 11.3 Table (s) 3.2-1 11.3-4 Figure Change (X1 Yes t 3 No Title of Change: Ring Water Vacuum Pump N6200C003 Changeout StM MRY:

This modification provided two new ring water vacuum pumps. One pump replaced the original nceth ring water vacuum pump. The original nortn pump wiu become a spare for the south ring water pump. The second new ring water pump is a spare for the new north ring water pump. The original Nash model H.4N is obsolete and no longer available. The replacement pump is a Nash model H-4.

The new pmp's peafcrmance characteristics are the same as those of the original pump. However, there are several minor differences between the two

models

1. The original pump design. fabrication, and material conformed to ASME III. ctees 3. The new pump meets the manufacturer's standards only. This is acceptable per Regulatory Guide 1.143,

" Design Guidance for Radioactive Waste Management Systems.

Structures, and Components Instau ed in Light Water Reactors".

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2. The news pump material, is cast iron whereas the original pump materiet is stainless stoet. Cast iron is acceptable for handling the domineralized water and gasses that the pump is expQced to.

3. The auction and afscharge flange size of tne new pump is sma W r. However, tha system ciosign pressure is less than the design pressure of the replacement pump and is therefore acceptabl.a.

4 The fun toad current is somenhat higher than the original, pump.

Howwver, the enlating cable and motor starter ratings are sufficient to handle the higher current.

Replacenent of the north ring water pump does not change the function, configuration, or operation of the offgas system. System rettability is maintatnad with the new ring wattse pump.

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I SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES Page 60 SAFETY EVAlliATION St&4AARY Safety Evaluation No: 91-0038 UFSAR Revision No. N/A Reference Document: EDP 8127 Section(s) N/A Table (s) N/A Figure Change I 1 Yee [X1 No Title of Changes Personnet Airlock Nandwheet Shaft Seal and Equalizing Valve Replacement SUW4ARY:

This modification replaced the personnel airlock handwheet teflon shaft seats with graphite seats. In addition, the 3" equalizing valves were replaced to eliminate the tefton seats and vetve stem seats. The new 2" equalizing valves use Nordet EPDM sesta and stem seats. The environmentatty quellfied graphite and Nordet components, having greater radiation resistance, are required to ensure containment integrity during tong term design basis accident radiation exposure in combination with calculated normal plant exposure levels.

I The difference in f rictional characteristics between the materials is within the design characteristics of the handwheet linkage assembly. The reduction in flow area due to the difference in the replacement valves' size does not affect ihe pressure equalization performance of the airlock. Therefore, the change in materials does not alter the function or operation of the personnet airlock and containment integrity is assured.

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1 SAFETY EVALUATIONS-ENGINEERING DESIGN PACKAGES Page 61 SAFETY EVALUATION StMAARY  !

Safety Evaluation No 91-0039 UFSAft Revision No. 5 Refarence Document: EDP 7121 Section(e) N/A Tabte(s) N/A Figure Change [X1 Yes [ ] No Title of Ohanges Permanent Installation of Condenser Sample Line SLMMRY This modification provided for the permanent instattation of a sample line connecting the condenser west outlet tap CT-N71-LOO 50 to sample sink #11. It replaced the tygon tubing Iris t atted by temporsey modification 88-0055 ' wit h stainless stoet tubing. The ' temporary modification was instaued to ensure that representative circulating water samples would be obtained.

Installation of the permanent s ample line ensures an accurate sample.

Accurate water samp1,es enst 's that the proper water treatment chemical dose rate is determined resulting in the biological organism control and the prevention of excessive chemical induced corrosion. Replacement of the tygon tubing with stainless stoet tubing allows the sample line to be returned to its original design basis. The sample itne routing. stress analysis, support design, and fabrication conform to the requirements of Detroit Edison Specifications and ANSI D31.1.

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SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES Page 62 FAFEW EVALUATION SUMRY Safety Evaluation No: 91-0044 UFSAR Revision No. 6 Feferonce Document: EDP 12234 Sectlon(e) N/A Table (s) N/A _

Figure Change (X) Yes ( ) No

-Title of Changer Seal Water Return Tank Sparger Installation and Vent Line Modification SU M Y:

The purpose of this modification is to prevent turbulence and tank pressure induced air carry-over from the seat water return tank to the condenser. This modification installed spargere in the seat return tank and increases the seat return tank vent size from 1/2" to 2". The spargers decrease the turbulence by dispersing the drain flow and the increased vent line size eliminates tank pressurization, j This modification enhances the operation of the seat return tank. It does not

! change the function of the seat return tank and does not interface with or challenge any saf ety related equipment or systems. This modification has been prepared in accordance with ANSI B31.1.0.

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SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES Page 63 SAFETY EVALUATION SLA4AARY Safety Evaluation No; 91-0047 UFSAR Revision No. _,,

5 Reference Document; EDP 12199 Section(s) N/A Table (o) 6.2-2; 6.2-13; 6.2-15 Figure Change (X1 Yes ( 1 No Title of Change: Division I Core Spray Minimum Flow /Rocirculation Isolation Valve E2150F031A Replacement SLA4WtY:

This modification replaced the Division I core spray minimum flow / recirculation isolation valve E2150F031A. The existing Limitorque operator was reinstatted on the new valve. A body test tap was also instatted as a field modification to allow Appendix J LLRT's. The replacement valve is the same make and modet as the original valve.

The replacement valve is identical to the original valve in terms of seismic qualification; code design and manufacture; materials of construction; actuation mode; and perf ormance - characteris tics. This modification does not change the performance and operation of the core spray Division I equipment and does not change the configuration of the plant.

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BAFETV EVAltla?!ONS EN0!NEERING DES!CN PACKAGES Page 64

$AFETY EVAlt1ATION CtM4MY i

t Safety Evaluation No 91 0051 REV 2 UFSAR Revision No. E s

Reference Documents EDP 12331 Section(s) N/A Tabte(s) N/A Figure Change IX) Yes [ ] No r

Tittu of Changes Station Air and Interruptible Controt Air Header Connections for the Future kotocation of the Instrument Air System (!AS)

Dryers and Receiver Tank OM AMY:

This modification added (1) a 3" connection, vatve, and cap on the 8" header downstream of the receiver tanks tocated on the first floor of the turbine bwit a'ing and (2) a 3" connection, valve. and cap on the 1AS supply to the Ranidua) Heat Removat (RHR) complex downstream of tne IAS 160tation vatve.

P5000F360. These air system tie-ine snow the on tine retocation of the IAS ,

dryers and receiver tank to make room for the instaustion of an additional du4ineraticer. The dominerttirer in scheduled to be instaued t

• ring the third rsfuoting e outage.

This modification does not impact any component or system ,c l a t ei. 4 ihe safe shutdown o f ' t he reactor. The connections are similar ic uis! tr g syst 4.

components and are instaued to B31.5 requirements.

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SAFE 7Y EVALUAY10NS ENGINEERING DESIGN PACKAGLd Page 66 SAFETY LVALUATION LtASAARY Safety Evaluation Not 91 0067 UFFAR Hevie ton No. 6 Reference Doctement CDP 12300 Section(s) N/A Tehte(s) N/A Figure C knge (X) Yes 1 1 No Title of Changet Hester Drain Pump Seat Water piping Pressure Dreakdown Orifice Instattation SLMMHY This modatication insteiled preature breakdown orifices in the beater drain pumps seat water intet piping. Three orifices are instetted in series to minimize cavitation and notes as the pressure is reduced. The orifices are staed to maintain adequate (2gpm) seat flow. This modification ieduces the heater drain pump mechanical seat cooting water pressure from 700 pet to the seat manuf acturer's reiommended pressure of 260 pet. This pressure reduction is expected to thcrease seat life end reduce the probability of seat failure.

The heater drain pumps are not required for safe shutdown, a cident feitigation, se for recovery after an accident. The addition of the orifices does not adversely affect the loss of feedwater accident analyses in the UFSAR.

SAFETY EVALUATIONS ENQ1N(( RING D&S!QN PACKAGES Page 66 f SAFETY EVALUATION Sl5BOMMY Safety Evaluation No 91 0059 UFSAR Revision No. _

5 Reforence Document: IDP 12301 Section(e) N/A Tabte(s) 6.2 23 G.7 15 Figure Change IX) Yes ( ) No Title of Chanos: Torus Water Management System ( TWS ) Outboard Con t ainn.an t Isotetton Vetve 06100F605 Replacement DUheERY:

This evaluation justifies the replacenant of the TYMS outboard isolation valve 05100F60$ with a new valve. The volve was reptoced because it could not meet the refurbishment reautremente for tocat teak rate testing seat tenkage integrity. The new vatve is the some make and modet as the former valve. The differences between the two vatves are materlat specifications and wedge type.

The new valve was forged and the old valve was cast. Some of the valve components in the new valve are made of different materials. The materlat substitutions meet Detroit Edison t'ermi 2 specifications. The wedge for the new valve is flexible whereas the originnt valve's wedge was notid, The flexible wedge represents an improvement and has no adverse affect on the performance of the valve.

This modification does not change the design bases, function, or operation of the TVNS. It does not have any affect on the oreration or reliability of interf acing plant systems. The new valve stroke time is slightly longer the originat vatvo stroke time. However. It is stitt within the existing stroke time design range.

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$AFE7Y EV84UAT10NS EN01NEER!NG DESIGN PACKAGE $

Page 67 SAFETY INAL11ATION SLMMRY Safety Evaluation No: 91 0064 UrfAR Revision N . t.

Reference Documents EDP 9168 Section(s) 9A.6.f.178_10.4 Tabte(s) N/A ,

Figure Ctange (X) Yes ( ) No Titto of Ctange . North Circulating Water Cooling Tower Modifications SLMMRY:

This modification incorporated miscotlaneous modifications to the north circulating water cooling tower to restore it to its originat condition and prc, vide the required cleculating water flow and temperature measurement points. This safety evetuation on\y addresses the modifications which af f ect the UFSAR. These changes include the installation of pitot tube isolation valves, insta Mation of temperature indicators, and replacement of the existing fi n with combustible f1M.

The circulating water system is not required for the safe shutdown the plant.

The new components do not interact with or support saf ety related equipment.

The instattation of combustible fitt does not invalidate the fire protection analysis because the circulating water cooling towers are located where a fire cannot affect any safety related equipment and the cooling tower basins are not used as a water supply for the ultimate heat sink or the - fire protection system. Installation of the pitot tube isolation vatves and temperature indicators have no impact on upon existing accident scenarios contained in the UFSAR.

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SAFE 7Y EVALUA?!ONS (NQ!NCERIN3 DES!GN PACVAGES Page 68 SAFETY [ VALUATION SUWARY Safety Evaluation Not 91 0073 UFSAR Revision No. 5 Reference Document: EDP 12428 Sectionte) N/A Table (s) N/A ,

figure Change IX) Yen ( ) No Title of Change Insta n ation of Fuel Pool Cooling and Cleanup System (FPCCS)

Hydrotesing Connection and Droin Line SUle%RYJ This modification added a hydro \asing connection and a drain line to the reactor wolt or dryer / separator storage pit to weste surge tank drain tire in the FPCCS. A tow spot in this piping located in the main hatlway of the second floor cf the reactor building has had very high radiation tevets (300-600 mr/he) due to plate out. This has caused the Division Il emergency equipment service water and residust heat removat radiation monitors to stay in the high alarm mode. The hydrolasing connections and drain line have been installed to clean the piping section so that radiation tevels can be kept to a minimum.

This modification also downgrades the modified piping section f rom Class C to Class O piping. The turbine building loop seat was filled with water to maintain secondary conta1hment integrity during the implementation of this modifica!!on. This toop seat w111 also be maintained during subsequent hydrolasing evolutions.

This modification does not adverseiy affact the FPeCS or any other plant system. When not in use, the hydrotesing connection will be bl,ind flanged and the drain valve win remain closed and capped. This modification does not affect secondary containment teskage and does not affect the ability of the standby gas treatment system to establish a negative pressure of 0.25" wg within 10 minutes after a LOCA. Administrative controls are in place to ensure that tuebine building loop seal is flooded to provide secondary containment int ege t *,y when hydrolasing is in progress. The piping downgrade is in ecspedance with Pegulatory Guide 1.26.

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SAFETY EVALUATIONS CNGINEERING DESIGN PACKAGET ,

Page 60 FNETY LYALUATION SLMAARY Safety Evaluation Not 9t-0110 UFSAR Revision No. ,_

6 Reference Documentt EDP.12749 Cection(s) N/A Tebte(s) N/A figure Change (X) Yes I 1 No Title of change Condensate Pressure Transmitter Instattation SUWMARY:

This change instetted a quick response pressure transmitter on the condensate side of the north number 3, 4 and 5 f eodanter heaters. This transmitter is used in conjunction wi'<h Generet Electric Transient Anstysis Recording System (CETARS) to record the transient condensate pressure response dueing reactor scrams. The pressure date will be used to investigate the cause of condensate thermet relief valve lifting problems experienced during reactor scrams. Yhe response of the normat condensate pressure transmitter is too slow for adequately recording of the pressure transients.

-The transmitter and tubing constructions material and inntattation is similar to the existing configuration, This modification does not affect safety estated plant systems. The condensate system and GETARS are not required to support equipment required for safe shutdown. Failure of the pressure transmitter tubing is taas severe than the outside containment feed

• ster line fatture anatyred in the UFSAR.

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SAFETY EVALUATIONS ENQ1kEERING DESIGN PACKAGES Page 70 WETY EV/LUATION Sa4AARY Safety Evaluation No 91 0111 UFSAR Revision No. 5 Ref erence liocunent: EDP t2770 Sectlon(s) N/A Tabte(s) N/A _

Figure Change (X) Yee  ! 1 No Title of Change: Relocation of Hester Drain Pump 01scharge Vent Connection SU>6%M:

This modification relocated the heater drain pump discharge vent connection for each heater drain pump from the heater drain pump discharge noarte to the discharge piping. The former threaded connection provided by the pump r manufacturer had failed on att three pumps resulting in steam teaks that contaminated the heater drain pump rooms. The new vent tines are connected to their respective heater drein pump discharge line with welded sockotet connectione. The new instattation reduces stress at the vent / discharge pipe connection resulting in a reduction in the likelihood of future connection fattures.

The new instattation meets the requirements of the ANSI B*1.1 Power Piping Code. It has no imonet on the function of the heater drain pump vents, heater drain pumps. instrumentation, or other heater drain pump system components.

The new vent connection tocations are at the same elevation as the former discharge notate vent connection locations. The new locations do not impact the discharge piping stress analysis.

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W ETY CVAlt%TIONS ENGINEERING DESIGN pACKACES Page 71 SAFETY EVALUATION !RMMRY Safety Evaluation Not 91-0114 UFfAR Revision No. 6 Reference Docunents EDP 17772 Section(s) 8.12 8,2 Table (s) N/A Figure Change (X) Yes ( ) No Title of Changer vain Transformer 2A Reptacement SLAM 4.RY This change replaced the original Ferranti main unit transformer 2A with an A00 trentformer. The original transforese was replaced due to an increase in combustible gas generetton within the transformer. This is indicative of transformer 011 breakdown due to localized overheating. The mounting configuration of the twe transformers is simiter. However, the ftem shunt bushing was modified to accommodate an approximate two inch gep between the tow voltage bushing and the isophase bus.

The replacsment transformer has the same MVA and voltage ratio rating, Engineering ceneiderations such as toad abaring and circulating current oateeen transf ortrors 2A and 20; voltage regulation; short circuit current 1 directior,a1 overcurrent relay settings 1 power feeder rating power uprate impact; and fire protection octuge system adequacy were reviewed and fourd acceptable. The operability of various compor.ents and ci*cuits associated with the main transformers was reviewed and either accepted or modifled to accommodate the change.

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SAFETY EVAlt1ATIONS ENQ1NEERING OESIGN PACKAGES Page 72 WETY EVALUATION S(MAARY Safety Evaluation Not _91-0120 UTEAR Revision No. 5 Reference Document CDP 12785 Section(e) N/A Tabte(s) N/A Figure Ctege (X) Yes ( ) No i

Title of Changet Elimination of separator Goat Tank North and South Outlet Level Controt Valve Bypass Valve Leaks SLM4ARY:

This modifications (1) removed the motor operatore on separator seat tank north and south outlet levet controt valve bypets valves, N2200F687 and N220F686 and (2) welded a plate to the valve neck of each valve to stiminate body to bonnet gasket teakage. Prior to this modification, on-line leak repairs were unsuccessful. Each ptete provides a more leakproof boundary and improves the integrity or the system. These valves were previounty abandoned by engineering design pockenes IDP 8938 and EDP 10778.

This modification does not change the function of the heater drains system or ariy equipment required for safe shutdown or accident mitigation because these valves were already abandoned. The plates were sized using the guidelines of ALME Section VI!!, Part 00-34 Equation 1. The plate instattations utilized full penetration welds to provide the necessary strength to maintain system integrity.

END OF EDP SECTION

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j FERMI 2 SAFETY EVALUATION

SUMMARY

REPORT 1991 PROCEDURES, TESTS, AND EXPERIMENTS

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SAFETV EVQLUQVIONS PROCEDURES. TLSTS AND EXPERifAENTS Page 1 SAF ETY EVALUAT1CH SUMMRY Safety Evaluation No ,00-0049 UFSAN Revision No. 6 _

Reference Documents g __90-074.UFS Section(s) 7.4 Table (s) N/A Figure Change ( 1 Yes IX) No Title of Change: Clarification of Reactor Core Injection Cooting (RCIC) and High Pressure Coolant Injection (HPCI) Suction Transfer Logic StMMRY:

This evaluation justifies revising UFSAR subsection 7.4.1.1.3.0 to state that a condensate storage tank (CST) fatture which results in a toss ct inventory and/or less of the current signal from either CST tevet transmitter witt cause en automatic RCIC/HPCI suction transfer. The original UFSAR tout stated that a complete failure of the CST and/or transmitter system would result in an automatic RCIC/HPCI suction transfer. Design Desis Task Force (DDTF) item E51 026 questioned whether dual upscale failures should be postulated that could disable the RCIC/HPCI suction transfer capability. Dust upscate failures are not credible because each transmittee system is redundant and the postulated failure would hatve to involve simultaneous failures of both transmitter systems. UFSAR subsection 7.4.1.1.3.8 was clarified to remove any implication that dvat upscale fattures of the CST tevet transmitters win result in a RCIC/HPCI suction transfer.

This change clarifies the description of the RCIC/HPCI suction transfer logic.

It does not atter the_the design, function, or operation of RCIC. HPCI. or the CST transmitters. The Formi 2 Saf ety Evaluation Report accurately describes the RCIC/HPCI suction transfee. It does not state that a complete failure of the CST transmitter systems causes a suction transfer.

SAFEVY EVALUAVIONS PROCEDURES. TESTS AND EXPERIMENTS Page 2 I

SAFETY EVALUATION St# ALARY Safety Evaluation No 90 0071 UFDAR Revision No. _

5 o

Ref erence Docunant s LCR 90-100.UFS Section(s) 6.63 g.13 15.6 Tabte(s) 9.1-1 ,

Figure change I 1 Yes (X) No Titte of Change Residual Heat Removat (RHR) and Core Spray (CS) Systems UFSAR Description Corrections SUheAARY:

This evaluation justifies correcting UFSAR discrepancies found by the Design Basis Task Force. These changes are as f oltons:

1) UFSAR subsection 5.5.7.3.2 has been revised to reflect the RHR system flushing methodology specified in system operating procedure (SOP) 73.205, " Residual Heat Removat Sys t em" . The UFSAR originalty stated that the RHR system was flushed through the minimum flow tine and the test line prior to initiating shutdown cooling. -SOP 23.205 speelfles that flushing is accomplished using the warm up line.

2) The RHR fust poot cleanup (FPCU) assist floarate in UFSAR subsection g.1.3.2 has been changed from the original 5400 gpm to 3500 ppm to reflect the actual flowrote achieved during startup testing and the results of a design calculation.

3) UFSAR subsection 15.6.5.5.1.1 has been revised to state that the RHR FPCU assist piping is seismic category I and that the non-seismic keep-fitt piping connects to the core spray piping. This subsection has stso been revised to ref erence UFSAR subsection 3.7.3.13 for the i design methodology for non-seismic piping that is connected to safety-related piping. The UFSAR originalty stated that the RHR FPCU assist piping is non-seismic and that there is no non-seismic piping connected to the core spray piping.

The plant has not been modified by these changes. The RHR FPCU cooling capacity still exceeds the maximum fust pool heat toad and the seismic design of systems that interface with the CS and RHR systems is in accordance with UFSAR section 3.7.

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SAFETV EVALUQVIONS PRDCEDURES. TESTS AND EXPERIMENTS Page 3

!AFETY EVALUATION StM M,RY Safety Evaluation No 90-0073 UFSAR Ruvision fea. 6 Refacence Documents ,1.CR 90 103.UFS Section(s) 5.6; 7.4 Tabte(s) N(A Figure Ctange ( I Yes (XI No Title of Change Reactor Core 1sotation Cooling (RC L ouction Tran6fer 1.ogic Description Correction SUh4RRY:

This evaluation justifies correcting the description of the RCIC suction transf er logic in UFSAR sut> sections 5.5.6.3.3 and 7.4.1.1.3.8. The description originatty stated that the au' >matic RCIC suction transf er to the torus on tow condensate storage tank (CST) tevet was single foiture proof. Design Basis Task Force item E51 027 questioned the accuracy of this statement aad identified components that would prevent RCIC suction transfer if they failed.

As a roe Tt, UFSAR sectioha 5.6.6.3.3 and 7.4.1.1.L 8 were revised to correcity state th.it the automatio tW suction transf er is accomplisher 1 by utilizing the

, de er eirgi z e to operate t*jp togic within the Division II HPCI system from redundant anatog CST te%i tran ritters and trip units.

tnt change makes UFSAR subsections 5.5.6.3.3 and 7.4.1.1.3.8 reflect the as-built condition of the plant. RCIC is not part of the emergency core cooling system (ECC5) and is not reqeired to achieve safe shutdown of the reactor. In addition, no credit is taxon for RCIC in accident mitigation.

Failure of the RCIC suction to align with the tor us on CST tow levet does not affect other equipment-involved in the accidents evaluated in the UFSAR.

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SAFETY EVALUMIONS l

PROCEDURES. TESTS AND EXPERIMENTS Page 4 SAFETY EVAil1AT10N SaWARY l l

Safety Evaluation Not 90-0001 UFFAR F.evision No. ,_5 l

DUTF B31 022 Reference Docw ont. thru 025-M __ Section(s) 6.6 Tabis(s) 5.5-1 Figure Change ( ) Yes IX) No Tin e of Changer Reactor Recirculation System (RRS) Pump UFSAR Text and Table Revisions StaWARY:

This evaluation just!fles revising portions of the UFSAR text and tables that coat with the RRS pump seats, seat cooling, motor, and required not positive suction head (NPSH) to address the concerns in Design Basis Task Force items 831 022-M through 025-M. The changes mee as fottows:

1. The operability lif e of the RRS pump seats has been changed from "1 year" to "1 opurating cycte" in UFSAR subsection 5.5.1.

2. UFSAR subsection $.5.1 has been clarified to specify that the RRS pump motoe is a standard AC induction motor capable of being operated with a power supply of varying frequency over the specified range and that the motor starts when the motor-generator excitation field breaker is closed.

3. UFSAR subsection 5.6.1 has been clarified to state that the reactor building component cooling water system supplies component cooling water to the RRS pumps via the divisionat emergency equipmer.t cooling water (EECW) system and that there are no RRS pump trips associated with the toss of RRS component cooling.

4 The RAS pump NPSH has been changed from 115 ft to 135 ft. This change reflects the octual operating point for hot conditions.

! These changes do net alter the RRS pump assembly system or components including their design bases criteria, function, operation, or controt. There is no change to the analyzed or postulated failure modes or offacts for the RRS pump assembly and its associated support systems and facilities.

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SAFETY EVALUAV!ONS PROCEDURES. TESTS AND EXPERIMENTS Page b SAFETY EVAlliATION SLMAARY Safety Evaluation No: 90-0039 REV 1 UFEAR Revision No. 5 ,

Pe terance Docuraent s LC't.90-005 tlFS Section(s) 3.51 10.2 Tabte(s) N/A Figure Change (X) Yes [ ] No Title of Change UFSAR Revisions Concerning Turbine Misaite Darriers SUWMRY:

This saf ety evetustion was written to address changes to the UFSAR concerning turbine missile barriers. During the development of a proposed licensing amendment to delete the turbine overspeed t echnical specification, inconsistencies were found ir the turbine mise 11e barrier support documentation f or severat UFSAR sections. These inconsistencies were evaluated in a design cateutstion. The design calculation confirms that the original design tevet of conservatism as currently stated in the UFSAR is maintained.

These changes to the UFSAR do not introduce a new mode of plant operation or involve a physical modification to the plant. No technicet specifications, assumptions, safety limits or timiting safety system setpoints are affected or changed.

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SAFE 7Y EVALUATIONQ PROCEDURES, TESTS AND EXPERIMENTS Page 6 CATETY 0 VALUATION fRMMRY Safety Evaluation No 9t'.0103 UFfAR Revision No. 6 Reference Document: LCR 90-134.UTS Section(s) 4.3; 10.43 16.17; 015.5 i Tabte(s) D16.1 1 )

I Figure Change ( ) Yes (X) No i

Title of Changet Revision to UFSAR Discussion of Core Power Distelbution and Cold Water Injection Effects SLAGAARY:

This evaluation justifies the following:

1. UFSAR subsection 4.3.2.5 has been revised to eliminate the statement that the core power distribution goat is that of a Hating shape and remove the implication that adjustments are needed when controt rode become - inoperable. The core power distribution is constrained by average planar lineer heat generation rate, tin 6ar heat generation rate, and critica*6 power ratio. The controt rod patterne are adjusted whenevee necessary to ensure conformance to the target exposure distribution and to thereat limite for a variety of reasons including but not limited to the presence of inoperable control rods.

2. UFSAR subsection B15.5.1 has been corrected to state that the cold water injection from an inadvertent high pressure coolant injection (HPCI) event results in a decrosse in intet enthalpy and a consequent increase in power. This original subsection incorrectly rtated that the above scenario resulted in a decrease in intet subcooling.

The changes in UFSAR subsection 4.3.2.5 de not result in changes to the facility or method of plant operation. Conservative bounds on the enount of axial, power distribution and restrictions on the manner by which the axial power- distribution may very are imposed. This distribution is monitored thr oughout the cycle within a pre-established licensing range. By staying within these limits, the consequences of any accident in UFSAR Chapter 15 are not increased.

The correction to UFSAR subsection B15.5.1 is an administrative clarification that only corrects the confusion between enthalpy and subcooling. It does not impset the function or operation of HPCI, has no effect on plant operation. and cannot affect the probability or consequences of any radiological, release.

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$AFETY EVALUATIONS PROCEDURES. TESTS AND EXPERIMENTS Page 7 SArF.TY EVAltiATION GlMMJtY F

90-0144 UFSA>1 Revision No.

Safety Evaluation No Section(s) 9.1 Reference Doctanent: LCR PO-208-UFS Table (s) N/A Figure Ctange (X1 Yes I 1 No Title of Changet New Fuel Assembly Transfer to the Spent Fuel Pool (SFP)

Storage Racks Using the Fuel Prepaastion Machine (FPM)

StM MRY:

The purpose of this safety evaluation is to justify using the FPM to place new fuel assemblies in the SFP stoaage racks, correcting therevising and description UFSAR offigure FPM operation in UFSAR sections 9.1.2.2.1 and 9.1.4.2.3, 9.1 8. The FPM is Soing used to lower the new fuel assembtles in the MP This is accomplished by storage rocks to avoid contaminating tbs erano.

that when a fuel assembly is placed in setting the FPM full, up end stop set so This ensures the FPM the bait ha.ndt e is just above the surface of the SFP.

that when a fust assembly is placed in the FPM the crane hook will not get wet.

UFSAR sections 9.1.2.2.1 and S.I 4.2.3 have been revised to eliminate reference to the use of used fuel channels (NRC commitment per DECO tetter #NRC-90-0078 dated 4/25/90) and to stateUFSAR thatfigure the FPM can be used for new fuel 9.1-8 has been replaced with a receipt / transfer activities.

simplified diagram of the FFW.

This change does not alter the fuel handling accident snalysis as the fust movement addressed in this safety evaluation is constrained to the SFP and the The consequerces of dropping a fuel fust assemblies are not irradiated. This assembly onto a SFP storage rack during pool transfer were evaluated.

change does not increase the likelihood of SFP storage rack damage because ests.btished administrative procedures require that the fuel assemblies be placed in and taken out of the SFP in areas which do not contain fuel storage racks.

The FPMs are not used to mitigste the effects of any accidents The or specificatty postulated to cause any accidents evaluated in the UFSAR.

possibility of inanvertent personnel radiation exposures by placing an irradiated fuel assembly in the FPM with the futt up end stop set to the new fuel receipt / transfer levet (tack of adequate snielding) la prevented by administrative controls.

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1 SAFET'r EVALUATIONS PROCEDURES. TESTS AND EXPERIMENTS Page 8 MFE1Y EVALUATION St& NARY Safety Ew tuation No: 90 0148 _

UFMR Rovision No. fi Reference Document LCR 90.2.15.UFS Section(s) 8.3 Tabte(s) 8.31 Figure ChanDe ( l Yes (X) No l

Title of Change: Deletion of UFSAR Cable Ampacity Table 8.3-1 SLAMARY:

The evaluation justifies: (1) Deleting UFSAR Table 8.3-1 which defines the ampacity for cables used at Fermi 2 and references to the table in UFSAR subsections 8.3.1.1.1 and 8.3.1.4.2.1 and, (2) Revising UFSAR subsection 8.3.1.4.2.1 to state that cable ampacities are con t roll,ed by design instruction. Table 8.3 1 is 11mited in its applicability in that it does not contain att cable sizes and types used at Fermt 2 and it contains no provisions for conditions that deviate from those upon which the table is based. The design instruction consolidates all ampacity calculations previously utilized at Fermi 2 including those from which UFSAR Table 8.3-1 was derived.

Both UFSAR Table 8.3 1 and the design instruction are based on the same critoria and industry standards. This revision does not change the plant.

Comparison of the ampacity values in the design instruction and UFSAR Table 8.3 1 verify that no existing cables are ef fected by these changes. Proper application of this design instruction practudes cable fatture, one of the potential causes of a single electrical component fatture, from occurring as a result of cable overheating.

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SAFETY EVQL%7ICWS PROCEDURES. TERTS A%) EXPlRIMENTS Page 9 CAFETY EVAutAfiOv4 rAsA%RY Safety Evaluation Not 90-0100_REV 1 UrSAR Revisinn 140 ,

6 Ref erence Doctament t LCR 90-719-UFS Section(s) N/A , _

Tabte(s) INA _

Figure Change IX) Yes ( l No Titt.e of Changet Control Center Pressure Boundary Shield Diock and Plank Wnu Modifications

!tA44ARY: -

This evaluation justifies revising UFSAR figuce 9.1-3. sheet 1 to rettoct the changes made by engineering design packags EDP-10125. This change made structural modifications to three *emovable sihield block and plank waM e on the fifth floor and one shield block and plank well on the third floor of the aux! Mary building. This modification upgraded the waM e from seismic category 11/1 to category I. The modification involved the addition of a continuous one eighth inch steel covee plate over the est\s in the standby gas treatment system rooms. The other shield plank waMs were structuraMy reinf orced with stoet plate and angles.

These watts are structuretty superior to the originat design. This modification compties with seismic category 1 design and implementing critoria. This ensures control contar pressure integrity in that the barrien are structureuy stable and capable of performing their safety function during a seismic event. The modification to the shield waM s does not change anything associated with the reactor building refusting floor ce its activities as discussed in UFSAR section 9.1.

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SAFETY EVALUATIONS PP.0CESURE S . TESTS AND CxPERIVENTS Page 10 l

IAFC1Y INAlt1ATION StAtAARY Safety Evaluation Not 90-0154 LTSAR Revisic,n No. 5 Ref erence Docunent COLR 3.0 Section(e) D.43 B.63 0.63 0.15 Table (s) D.4.3-1: 0.5.1-13 0.6.1-1 D.15.0-13 B.15.0-2: D.15.0-3 D .15.1. 2 - 1 3 0.15.1.2-2:

D.15.1.3-3: 3.15.1.2-4:

D.15.1.2-5: B.15.1.3-1:

D.16.1.3-2: 0 15.1.3-3 D.15.1.3-4 _

Figure Change ( ) Yes (XI No Title of Change Revision to Core Operating Limite Report SLN MRY:

This evaluation justifies changes to the Core Operating Limits Report (COLR).

These changes include:

1. Changing the methodologies f or calculating thermal limits as a result of acquiring average power range monitor and rod block monitor technical specifications (ARTS) and Maximum Extended Operating Domain (MEOD). The mamimum everage linear heat generation rate (MAPLHGR) and minimum critical power ratio (MCpR) have become a function of reactor power and core flow. As a result of MEOD. the rod block moniter (REM) trip setpoints and allowable values have become cycle specific and have been added to COLR-3.0.

2. Adding the thermal limits for the new fuel types used in cycle 3.

228 GE6 fuel bundles were replaced with 224 GE9D fuel bundles and 4 SVEA-96 toad fuel assemblies (LFA). MPLHCR and and '.4near heat generation rate (LHCR; limits for the fuel bundles loaded for the first and second cycles remain the same.

3. Using 00 mil f uel channels on the new GE90 fuel bundles The fust bundles loaded during the first and second fuel cycles were enclosed in 100 mil fuel channels.

The addition of the MPLHGR. LGHR. and operating MCpR limita to COLR doss not change the method of plant operation or result in any modifications to the facility.

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SAFEVY EVALUATIONS <

PRZEDURES, TESTS AND EXPERWENTS Page 11 l

l Safety Evaluation No. 90-0154 (continued):

Use of the GE9B fust design has been genericstly approved by the NRC. Its application at Fermi 2 has been analyzed using the approved methodologien in GESTAR II.

The perf ormance and function of the SVEA.96 LFAs are similar to that of the GE90 reference bundte. They comply with the criteria in GESTAR II. Use of tre LFAs does not changs the operation of the plant or require modifications to equipment important to safety. The LFAs are geometricatty compettbte with existing fuel assembtles, control rods, nautron instrumentation, spent fuel storage - rocks, and f ust handling equipment. The GE cycle 3 reload analysis ,

using the GE98 reference oundle remains valid because (1) the shutdown margin is slightly improved with the SVEA-96 LFAsi (2) the peak cladding temperatures are 60 to 160 degrees F tower than an 8x8 bundle operating at the same LHGR and i

maximum oxidation rates are far below the 17% design timits and (3) the differences in void coefficient, fuel rod thermat time constant, and scram reactivity win not effect the overpressurization transient because of the smatt number of LFAs in the core.

The purpose and performance of the 80 mit channets are identical to that of the 100 mit channets. The 80 mil channels do not require any modification to equipment important to saf ety and do not change the method of operation. The 80 mil, channets are geometricany compatible with existing fuel assembtles, controt rods, neutron instrumentation, spent fust storage racks, and fust handling equipment.

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SAFETY EVALUATIONS PROCEDURES, TESTS AND EAPERIMENTS Page 12 LAFETY LVALUATION SLMM,RY Safety Evaluation Not 91*0001 UF!%R Revision No. 5 Referonce Document: LCR 91 001-UFS Bection(e) 9_ .1 Table (s) N/A Figure Change ( 1 Yes IX] No Titte of Changes Fuet Assembly Drop Height Over the Spent Fuel Poot Revision to UFSAR SlMMRY:

This revision changes the fust assembly drop height in UFSAR section 9.1,2.1,1 for a drop in the spent fuel pool from 6 ft b an. to 17 ft 7 in. Detroit Edison re-evetusted the integrity of the opent fust poot liner for a postulated f ust assembly drop of - 17 ft 7 in.3 the maximum height currently permitted by the ref usting bridge. The results of the evaluation indicate that the spent fust poot liner is capable of sustainin0 a fust assembly drop from a height of 17 ft 7 in. without penetrating the liner. These results are documented in a design calculation.

No new design is required and no modifications were made as a result of this change. The spent fuel poot f ust assembly drop accident discussed in UFSAR Chapter 9 is unaffected by this change because the spent fust pool linee integrity is still maintained. The radiological consequences of the spent fust pool fuel assembly drop are not affected because they are bounded by the fust assembly drop over reactor core accident.

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SAFETY EVALUATIONQ PROCEDURES TESTS AND EXPERIMENTS Pege 13 4

i SAFETY EVALLIATION SLMARY l Safety Evaluation Not 91 0006 UFSAR Revision No. &

Reference Documents LCR 91 010.UFS Section(s) 9A.6 _

Table (s) N/A Figure Change ( ) Yes IX1 No Title of Change: Clarification of SuevoiM ance Requirements for C0 Fire Protection System Valve Verification SUh&MHY:

This evaluation justifies adding a clarification note to UFSAR section 9A.6,4.2 which states that the vatvo tineup for the CO fire protection system is verified by verifying that tre CO, tank levet )and pressure are within the acceptance criteria of the applicable surveluence taste. This m6thodology is required because the position of these valves cannot be determinea by visual inspection. This clarification was previously contained in Technical Specification Clarification TLC 89 022. However, the fire protection requirements have been removed from the technical specifications and placed in the UFSAR. As a result, the clarification has been incorporated into the UFSAR.

This change is a document clarification and no physical changes were made to any component or system in the plant. This clarification does not impact any survoittance or operating procedures.

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fM EVY EVAt.UAV10NS PROCEDUMES, TESTS Ado LxPER!ufHit Page 14 SAFETY EVALUATION CAWRY Safety Evaluation No 91-0012 Uf'SAR Revision No. 6 Reference Documents: 67.000.601 Section(s) N/A 78.000.69 Tabte(s) N/A Figure Change IX) Yes ( ) No Title of ChanDe Alternete Method of Grab Sampling the Primary Containment Atmosphere

$lhedARYt This evaluation justifies the use of an alternate method for grab sampling the primary containment atmosphere. The atternate greb sampting scheme draws a sample from the drywett through the Division i H2/02 cabinet H21P282 sample tine into a manual sample rig. The sample is returned through the Division !

H2/02 Cabinet H21P282 sample return line. This sempting method violates containment integrity when samples are being taken because the sample suction containment isolation valves are remote manual isolation valves and do not automaticetty isolate on a containment isolation. In addition, the manuat sample rig is not expected to withstand the theorized primary containment pressures and temperatures that would suist foM owing a LOCA. The normat method of sampling the primary containment does not violate primary containment integrity because the sample taps are tocated outside the primary containment boundary and isoleted by the primary containment radiation monitoring skid (PCRMS) automatic isolation valves. If one of the four PCRMS automatic isolation val.ve s has to be doenergized c'.ased or faits closed, the normat sampling method introduces the risk of not meeting the required sampling frequency f or drywett purge and venting. The inability to draw grab samples could require a plant shutdown if the PCRMS automatic isolation valves are not rostored to operability within the Technicat Specification time limits.

In order to meet the functional requirements of primary containment, the atternate method of grab sampling requires the following administrative controtes

1. The teakage area of the alternate sampling tapa has been reduced by instau ing 0.25 in, diameter fittings.

2. A method of providing primary containment isolation witlin the normat containment isolation time frame is used white alternate grab samples are being taken. This method dedicates personnet to monitoring control conter instrumentation for LOCA conditions and to deenergizing the power supply to the alternate sampling remote manual valves when LOCA conditions are detected.

SAFE 7Y CWLtMIONS PROCEDURES, TE&tS AND EXFERIMENTS Page 15 i

Safety Evaluation No. 91-0012 (continued):

l 3. The atternate grat, sample taps are controtted as locked valves.

1 The grab sample tap locations were selected auch that their primary A. l centainment isolation valves are the same modet as the PCPMS sutomatic f' isotation volves.

I There is no loss in the tevet of protection provided by any of the fission  !

product barriers. The administratfve controte for the atternate sampling ,

method provide the capability of isolating the e smp1.e lines within the 60 )

second primary containment isolett. Ame. An analysis of the frequency of use and the probability that the administrative controls would be used concluded

-that the - probability of a LOCA occuraing white alternate sampling is in progrees is-. below that of -nortnet regulatory . concern. Therefore. the probability of UFSAR/SER st.atyted accidents and transients is unaffacted.

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GAFF;7Y EVALUA?a N9 PF.CCEDUARS, TESTL AW I!XpERIMENTS 3 Page 10 i

FN ETY EVALt1ATION e M RY Safety Evaluation Not 91-0016 REV 1 UFMR Revision No. N/A Heference Douument: DER 88 2071 Sectlon(s) N/A _ _

Tabte(s) N/A Figure Change ( ) Yoo (X) No Title of Change: Temporary Storage of Mixed Weste in the Onsite Storage Facility (OSSF)

SLs44W:

The purpose of this ovatuation is to justify the temporavy storage of mixed weste at the OSSF. The mixed waste enneists of freon (2% by volume) contaminated waste body o11 generated as a resdt of dry cleaning laundry operations. The mixed wastu is stored in DOT-17E $6 gaMan steet drums. The  !

storage area is located in the northeast section 3f tha OSSF and occupies an area 16 ft by 30 ft. A containment berm is instatted in the storage aren to prevent mixed weste from en'ering the floor drain system in the event of a storage drum rupture. Mixed waste is being stored at the OSSF because it is currentty not accepted at low tevet radioactivo disposet sites or hazardous we.ete alsposat facitities. This is a terrporney condition that win test until, regulations attow the weste to be processed into separate hazardous and radioactive constituents.

I The mixed waste le non-emptosive and essentiaMy non-votati'le. Siting the mixed waste storage facitity in the 0$3F wl M result in radiation exposure consistent with Al. ARA guidelinen. The isotopic concentrations average less than 10% MPC and dose rates on the outside of the d*ums- are essentleMy nonexistent.

Storage of mixed waste at the OSSF wi u not increase the radioactivity inventory of the radmaste system. The OSSF is designed to prevent the release of radioactive materiets to the environment and meets the requirements of the Assnurce Conservation and Recovery Act contained in 40CFR266 f or the hazardous weste components of - the mind waste. The storage drums are designed to Department of Transportation DOT 17E specifications. Therefore, no severe drum damage due to a seismic event ne faM should occur. Any release of mixed waste within the OSSF will be contained by the berm. The applicabte portions of Aegutatory guides 8.8 and 1.143: NUREG-0800; IE Circular 80 18 10CFR20.206:

10CFR60.34A;.and 10CFR60 Criteria 60. 63, and 64 were examined and the storage of mixed waste in the OSSF was found to be acceptable.

4

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SAFt,VY EVALUATIONG PROCEDURES. TESTS AND EXPERIMENTS Fege 17 CAFETY EVALt1AT10N SLAOMRY Safety Evaluation No 91 0018 UFFAR Revision No, b Reference Document: LCR 91 051-UFS Section(s) 9A.4.7 Tabte(e) N/A Figure Change ! ] Yes (X) No Title of Change: Division 1 and !! Underground Safety Related Ducts Fire Harerds Analysis SLA4M,RY:

This evaluation justifies (1) a revision to the UFSAR fire hazard analysis of equipment and structures in tfie yard area to address the Division I and II underground safety related cable ducts between the residuet heat removat (RHR) complex and the auxittery building (AB) cable vaults and (2) corrective action for a CA audit finding generated during the September, 1990 fire protection trienniet audit. The QA audit finding tridicated that flanmebte 11gulds (f rom a tank truck oil spilt fire) mey pose a threat to the Divissen 1 and Il saf ety related cable ducts rutining between the RHR complex and the AD. The ccncern was that the burning liquid would enter both division manhole covers and damage the safe shutdown cables in both divisions. The cable ducts contain safe shutdown cables for the RHR service water system, emergency equipment service water system, and the emergency dieset generators. At the time of the finding, one manhole was complesely covered with soit and gravet. It was determined that the as-found soit and gravet cover and the physical separation of the division cable ducts had provided adequate protection to ensure that ons division was available for safg- shutdown during and after a fire. The audit finding corrective action includedi (1) covering the manhole covers with soit and gravet to provide a physical barrier against the flammable 11guld and heats (2) erecting a rope barrier and signs to control the storage of combustibles on er between the manhotest and (3) inspecting annually to assure that the rope barrise and-signs are in place.

The fire hazards analysis used the guidance included in the NRC's response to question 3.1.4 of Generic let ter 86-10. The analyste concluded that at '.e a s t one safe shutdown train remains available during and af ter a fire. Tne soll and gravet overlay and barrier are sogareted from the safe shutdown cirt.uits.

The eddition of soit and gravet over the manhole covers reduces the amount of fire damage to the safe shutd;wn cables by functioning as a flame arrester.

The probability of a fire within the area of concern is significantly redLeod becaus6 storage is not permitted in the area. The corrective actions have an insignificant effect on seismic, fl.ooding. and tornado design.

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SAFETY EVALUAT!EAS PROCEDURES, TESTS AND EXPER1M NTS Page 18 SAFETY EVALUATION SLBMARY Safety Evaluation No 91 0019 UFSAR Revision No. $

fief erence Document : 82.000.18 Gection(s) 9.1 Tabte(s) N/A Figure Chan0e ( 1 Yes IX) No Title of Changes Additionat Control Rod Storage Rod Storege in the Spent Fuet

,8 cot (SFP)

St.ANARY This evaluation justifies the addition of 40 control rod curb hangers in the SFp to provide enough storage for the control rods replaced during .he second refueling outage. The hangers ute designed to be mounted on the f our inch high curb around the $FP. Each hanger has two lifting / storage rode of different lengths and each lif ting / storage rod supports one controt rod. The staggered hanger rods maximize the use of the available space in the SFP. ihe lifting / storage rods **ve a welded tab for position indication purposes. With the reactor building crane auxiliary hoist in its fuit up position, the tab is set to the top of the SFP curb using a chainfatt. A lock is attached to the chainfall to prohibit f urt her ' upward travet of an engaged control rod. This ensures that the contret rod is maintained at greater than 6 ft 6 in, below the ref usting put f arm tr acke as required by Technical Spect f Acations.

The curb hangers store the control rods at a tower elevation than the highest instan ed wa n books. Seismic and structuret load analysee indicate that a curb hanger with a control rod a t tached will withstand a design basis earthquake.

The fust handling accident analysis does not address a control rod drop $n the spent fuel pool, l'oweve r , the impact energy from a postulated controt rod drop is less than that anatyred *or a fuel assembly drop. Administrative controts and hoist bl ock s ensure that the 6 ft 6 in. minimum height restriction of Technicat Specificatirn 3.9.6 is met. The load limit for crane travet over the SFP stated in Technical Specification 3.9.7 is 1100 lbs greater than the combined weight of a control cod, curb hanger, and chainfall.

l

SAFETY EVALUAVIONS PROCEDURES, TESTS AND EXPERIMENTS page 19 5AFETY EVALUATION SUh4MRY Safety Evaluation No: 91 0026 UFSAR Revision No. N/A Reference Documont: SOE 91-01 Section(e) N/A .

Table (s) N/A Figure Change [ ] Yes (X) No Tit' e of Changes Simulation of Drywen Gas Sampling Techniques Using Sequence of Events proceduce SOE 91-01 SLWhMRY:

This evaluation supported the SOE 91 01 procedure for mass spectrometer testing of the primary containment atmosphere (PCA) grab sampling technique used by radiation protection from 9/88 to 9/90. It was believed that this sampling technique introduced air from the reactor building atmosphere into the sampling system which diluted the semples. It was also believed that a back-flow condition caused by the intmakage stopped flow through the oxygen and hydrugen sensors. As a consequence, it was t hought possible that the primary containment monitoring system (PCMS) and, possibly, the primary containment radiation monitoring system (PCRMS) vare inoperable during the duration of the sampling intervat. SOE 91-01 simulated the sampling technique and traced the air inteakage flowpath through the sampling system. The test data was used to cetermine PCMS r.ce.'abili t y , PCRMS operability, and whether representative sampling of the PCA prior to venting or purging the drywe n was occurring.

The performance of this test m limited to operational conditions 4 and 6. In operational conditions / 4hc: the PCMS and pCRMS are not required for acetdents previously eve '.c% r e > lo she UFIAR. This test made no structurat design changes to the PCMS or PCRMS u described in the UFSAR.- Helium was used as the tracer gas. Its concentration in primary containment was negligible and it is not corrosive to primary containment or to the equipment within containment.

m -. .

SAFETY EVALUATIONS PROCEDURES. Ti.STS AND EXPER1W.r4TS Page 20 SAFETY EVALUATION SLMMRY Safety Evaluation No: 91-0030 UFSAR Revision No. 5 Reference Document: LCR 90-068.UFS Section(s) 6.3 Table (s) N/A Figure Change t 3 Yes IX) No Title of Change Clarification of UFSAR Description for Emergeacy Core Cooling System (ECCS) Discharge Line Filt System Low pressure Alarms SLMAARY:

UFSAR section 6.3.2.2.5 has been clarified to specify that fitt system low r *sure alarms exist for the low pressure coolant injection (LPCII and Lore us fCS) systems. The original description stated that the high pressure cv injection system (HPCI) also had a flu system low pressure alarm, frr e system cischarge piping is kept fun by the head of water from the um to storage tank (CST) and relies on the CST tow levet alarms for m "o w sf an inadequately fitted discharge line.

S.a .nge does not create any physical mo.lificatione to the plant and does not cnance any operating procedures. The function and operation of the ECCS discharge line fill system is not changed by this clorification. Operability -

of the HPCI system is assured by the physical design of the HPCI fl u system in which water and pressure is supplied by the head of the CST. The requirements provided in the basis for the Technical Specifications. UFSAR. and Safety Evaluation Report are not affected by this change.

SAFETY EVALUATIONS PROCEDURES, TESTS AND EXPERIME.NTS Page 21 SAFETY EVALUATION St#4MRY Safaty Evaluatlot No 91-0031 UFSAR Revieion No. N/A Reference Document: N/A Seeflon(s) N/A Table (e) N/A Figure Change i 1 Yes IX1 No Title of Changes Processing Chemical Weste Tank (CWT) Contents Through the Condensets Phase Separators (CPS)

SLM RRY:

This evaluation justifies processing the contents of the CWT in the CPSs. This method consists of transferring 2000 gations or less of neutretired chemical waste to a CPS containing 100 cubic feet or more of partiaMy expended ion exchange resins. The phase separator contents is circulated for one hour minimum. Contact between the resins and choolcat waste results in the removat of ionic and organic impur.t..es. The supernatant contains less than detectable radionuclides and a significant reduction in total organic carbon (TOC) is achieved, Normany, the contents of the CWT are processed by the foMowing

[

t methods:

1. The CWT contents are transfe. red to the floor drain coMector tank (FDCT) and subsequently processed through the floor drain fitter (FDF), evaporator feed surge tank (EFST). and floor drain and weste domineralizers (F00 & WD) for release from the waste sample tanks-(WST).

I

2. The CWT contents are processed through a portable activated carbon bed l

and mixed bed domineraliser. The affluent is transferred to the FDCT and processed as described above.

3. The CWT contente is: transferred *.o-the FDCT and processed through the waste collector fitter for release from the WSTs.

I Method 1 creates additional solid radweste and an unacceptable TOC in the WSTs method 2 is cost prohibitive.: and me t hod . 3 is contrary to the goats of the Fermi 2 zero discharge program. Procersing the CWT contents in the CPS supports the zero discharge and solid radweste reduction programs.

SAFETY EVA,' VATIONS -

PROCEDURES, TESTS AND EXPERIMENh, Page 22 Safety Evaluation No, 91-0031 (continued):

No equipment important to safety is used to process chemical waste and no equipment. important to safety is located in the redweste building. The chemical, waste has no effect on the radwaste system components because it is neutralized prior to transfer to the CPSs. Since the chemical waste is-processed in the redweste system prior to release to the WSTs. there is no effect or, releases - to unrestricted areas. This method does not affect the Formi 2 process control program. The calculated offsite doses that would result due to the use of this process and as a result of a hypotheticat redweste accident are greater than those previously calculated in UFSAR section 15.7.3 but are atitt much less than the limits in Appendix B of 10 CFR20.

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SATETY EVALUATIONS PROCEDURES, TESTS AND EXPERIMENTS Page 23 fATETY EVALUATION SUBAMRY Safety Evaluation Not 91-0035 UFSAR Revision No. _

5 Reference Document LCR 91 079.UF9 Sec t ior.( s ) 7.1; 7.6 Table (s) N/A Figure Change ( 1 Yes (X1 No Title of Change Revising the T50 Primary Containment Radiation Monitoring System (PCRMS) Functional Description in UFSAR Sections 7.1 and 7.6 St ANAHY; this evaluatin justifies revising UFSAR subsectiono 7.1.2.1.22.1(A) and 7.6.1.12.1.6 to make the T50 PCRMS functional description consistent with UFSAR Table 6.2-16; subsections 7.3.2.2.7.1 and 7.3.2.2.7.68 and T ec hnic at Specification 3.4.3.1. The new wording eliminates any discussion that states that the T50 PCRMS is required: (1) when the reactor is shut down (i.e., cold shutdown or refusting), (2) when personnet enter containment, or (3) when the atendby gas treatment system (SGTS) operation is esquired. This revision clearly states that the T50 PCRMS functions as part of the leakage detection system (LDS) durin.g operational conditions 1,2 and 3 enty.

This revision does not change the function of the TSO PCRMS as it is currently utilized at Fermi 2. The criticality radiation monitors fulfill the criteria of NUREG 8.12 (UFliAR Appendix A.8.12) for a refusting droppao rod accident. In case of an accidont, these monitors are supplemented by the control center normat makeup air radiation monitor and the control conter direct radiation monitor to fulfill General Design Criteria 13. The Fermi 2 Radiation

-Protaction Program as established in UFSAR Section 11.9 f ulfille the radiation protection requirecents set forth in 10CFR 19 and 20s the applicable regulatory guides; and the Tochnical Specifications. Administrative controls, portable radiation monitorirg instrumentation, and the area radiation monitoring system ensure that personnet exposure is kept as low as reasonably achievebte (ALARA). Therefore, this change does not increase the the onsite or offsite dose received by personnet f onowing an accident. The PCRMS is not quotified to operate under post accident conditions. The containment area high range monitors perform primary con t a inmer,t monitoring during and after en accident

! and the SOTS offluent SPINGS are used to monitor post accident offl,uunts, i

I L

1 SAFE 7V EVALUAVIONQ PROCEDURES, TESTS AND EXPERIMENTS Page 24 EAFETY FVAltaATION StMMRY Safety Evaluation No 91-0041 _

UFSAR Revision No. N/A Re f erence Doctanent : DER 91 0102 Cection(s) N/A Table (s) H/A Figure Change i 1 Yes IX) No Title of Change: Control Center Heating, Ventitation, and Air Conditioning (CCHVAC) Ductwork Leak Test Acceptance Criteria SLM MRY:

This evaluation justifies modif ying the acceptance criteria fcr teak testing the CCHVAC ductwork based on revised assumptions for unfittered intenkage end for roouced ingress and egress inteakage from the present control conter door configuration. Roanalysis of the control conter personnet exposure received in radiological emergencies provides the basis for changing the acceptance criterie. The ottowabl,e personnet dose is 30 rem to the thyroid over a 30 day period in accordance with General Design Criterion 19 and Stancard Review Plan NUREG-0800. The originat dose assumptions assumed a constant unfittered control center inflitration rate of 10 cfm.over the entire 30 day period, 'The original calculations resulted in a thyroid dose to control center personnet of 16.1 rem, The new assumptions assume an unfittered controt center infiltration rate of 32 cfm for-the first 30 minutes due to a postulated damper failure in the operating CCHVAC division and subsequent switch-over to the other division within 30 minutes; an unfittered control canter infiltration rate of 9 cfm for the balance of the 30 days; and a constant 3 cfm for operating personnet ingress and egress over the entire 30 day period. The results. of the remnalysis indicate that the thyroid dose to control center personnel would be

'18.7 rem.

l The CCHVAC system will still provide an acceptable post-accident environment with the new duct and door intenkage assumptions. The calculated thyroid done

(. to control center personnet is well below the 30 rom limit. Changing the teakage criteria does not change the design or function of the control center or CCHVAC.

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SAFE 7Y EVALUATIONS PROCEDURES, TESTS AND EXPERIMENTS Page 26 7.ATETY EVALUATION SLBAWRY Safety Evaluation No: 91-0053 UFSAR Revision No. N/A Reference Document: DER 91-0330 Section(s) N/A Table (n)- N/A Figure Change ! l Yes IX1 No Title of Changet Justification for Using the Damaged Refueling Bridge Mast During the Second Refusting Outage SLAAMRY:

This evaluation justifies using the refueling bridge mast during the second refueling outage with a horizontal bracing member on the outermost mast truss removed. This structural rember was damaged during load testing when the mast assembly was raised too high. It was removed f or engineering observation and evaluation. The structure was returned to its original configuration after the second refusting outage.

The structural configuration of the refucting bridge mast during the second refuating was acceptable for the following reasons:

1. The forces acting on the outermost mest truss are due to the weight of the mast and the drag generated as the mest in being moved through the water. The weight of the mast is carried by the vertical truss tube sections.. The horizontal drag forces are unif ormly distributed over the length-of t he ma s t and, as such. .the load is redistributed over the remaining structural members. The adjacent horizontal and diagonal members easily compensate for the 1,o s s of the horizontal member. The connection between the grapple and ball is not considered to be rigid. Therefore, there are no significant bending moments escried by the meet structure.

2. A confirmatory test was performed to determine acceptable operation of the mast. A dummy fuel, bundle was lifted out of the spent fuel pool, placed in the reactor cavity, and returned to its original position in the spent fuel pool. The acceptance criteria of this test verified that the truss sections move up and down without binding and that the grapple satisf actorily perf orms its tatching and untatching function as intended.

i

SAFETY EVALUATIONS PROCEDURES. TESTS AND EXPERIMENTS Page 26 Safety Evaluation No. 91-0053 (continued):

The mast is only used as a cable guide. The UFSAR does not describe an accident dealing directly with the fuel mast, Other accidents pertaining to fuel storage and handling described in the UFSAR are not affected by the structural configuration described here. The missing most menber did not adversely affect the function of the grapple. Use of the mast in this configuration did not affect the operating condition of the rof uel.ing pla t f orei; hoists or cranes for handling control rods and f. sol assemblies; core internats; or the pressure vessel.

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. SASETV EVALUATIONS PROCEDUHES, TESTS #4D EXPERIMENTS Pape 27 SAFLTY EVALUATION SLSAMMY Safety Evaluation Hos 91-0058 UfGAR Revision No. _N/A ,

Reference Document: COLR 3.1 Section(e) N/A Table (s) N/A Figere Change ( ) Yes IX1 No

'. " e of Changer

, Revision to Core Operating Limits Report (COLR)

SLASMJtY:

This evaluation justifies the fo nowing changes described in COLR 3.1:

1. The maximum everage planar linear heat generation rate (MAPLHCR) and the linear heat generation rate (LHGR) 11mits for the SVEA-96 toad

' fuel assembtles (LFA) have been revised to accommodate a change to the process computer model. The process computer database assumes the LFAs have 60 f ust pins. This is the same number of pins that GE9B fuel, bundles have. The LFAs. have 96 pins and, as a result, the calculated heat generation rates are 60% higher than the actual values. To anow for the proper calculation of the thermal limit margins, the LMPLHGR and LCHR timits have been raised 60L

2. Rod bl.ock monitor ' (ROM) f Rter time constants and RDM setpaints have been revised. Specific RBM fitter time constants have been added to ensure the ef f ectiveness of the filter is not degraded. Yhe RBM ticonsing basis Supports any combination of time delays and fitter

+ime constants that are anowed by the system hardware. However, some combinations reduce the filtere effectiveness. An ineffective filter, combined with the more limiting setpoints required to use it, results in more f requent unnecessary rod blocks than en unf t?.tered system.

The RDM setpoint aMowable values have been adjusted to correct a smatt variance betwown the COLR 3.0 values and the values based on a new Fermi 2 - specific design cateute t ton. The original RBM netpoints were obtained using a generic analysis whirh is applicable to Formi 2.

The analyticet timite and nominal trip setpointo are the same in both the calculation and the ' generic analysis. However, the attowable values for the power and trip setpoints are stichtly different.

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k SAFETY EVALUATIONS PROCEDURES.-TESTS AND EXPERIMENTS Page 28 Safety Evaluation No. 91-0058 (continued):

Thermat timits cateutations have been performed using the NRC approved methods in GESTAR !!. The thermat limit changes do not invalidate the reload ticonsing analysis. Revision of the LFA thermat limite does not result in a modification to the facility and does not change plant operation.

The filter time constants, and attowable setpoint values specified in cat? 3.1 do not invalidate the rod withdrawat error analysis and have no effect on any other transient. The specifications in COLR 3.1 do not result in any modification to the facility.

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SAFEVY EVALUAV10NS

' PROCEDURES. TESTS AND EXPERIMENTS Page 29 SAFETY EVALUAT10 St# MARY Safety Evaluation Hot 91-00G1 UFSAR Revision No. 5 Reference Document LCR 91-089-UFS Section(s) DA4.1 DA4.23 9A4.10 Tabte(s)

Figure Change (X) Yes ( ) No Title of Change: Control Room Fire Hazards Analysis

SUMMARY

This evaluation justifies expanding the discussion an the UFSAR hazards analysis for the controt room and reactor building fifth floor fire zone. This discussion includes the air conditioning room on elevation 677'-6" of the aux!Liary building in the control room fire area and the auxiliary building stairwett (columns F9 to Fl *2 ) in the reactor building fifth floor fire area.

It also provides justification for having no fire detection instrumentation in these rooms. The Appendix R safe shutdown analysis considers the air conditioning room as part of the control room fire area and the auxittery building stairwell as part of the reactor building fifth floor fire zone. A justification of the lack of fire protection in these two roome is required because they contain safety-related cables.

The air conditioning room does not contain any cables or components requi.'ed for the safe shutdown of the plant. This room does not contain any exposed combustible materiets as att combustible materlate are encased in metal housings (motor housings, cables in condult). The room is heavity congested with~ conduit precluding the future storage of combustibles. As a result, no fire is postulated in the air conditioning room.

The auxiliary building stairwell contains two Division 11 reactor and auxiliary building HVAC system cables which are required to achieve and maintain safe shutdown. The safe shutdown analysic documents the fact that the toss of these cables due to a postulated fire would not prevent the safe shutdown of the plant because of the availability of Division I systems outside of the fire area. All combustible materlats in the room are oncesed in metal enclosures (a gang box, cables in condult) and are not readily accessible. Plant procedures do not attow storage of combustibles in stairwells. As a s'eault, no fire is postulated in the stairwelt.

This revision to the UFSAR fire hazards analysis does not affect the conclusions of the fire hazards analysis or the Appendix R safe shutdown analysis. There are no physical plant changes made by these changes.

SAFETY EVALUATIONS PROCEDURES. TESTS AND i3pER1MENTS

.Page 30 SAFETY EVALUATION SthmRY Safety Evatuation No: 91 0065 UFCAR Revision No. 5 Reference Document: DER 91-0428 Section(s) A.1.52 Table (s) N/A Figure Change i 1 Yee (X) No Title of Change: Control Center Fil,tration System Charcoat Adsorber Testing Requirements SUW WRY:

The purcose of this evaluation is to justify a revision to the UFSAR in which compliance with Regulatory Guide 1.62 is modified to reference the requirements of Tabte 5.1 of ANSI /ASME NSO9-1980. The requirements of Table 5.1 of ANSI /ASME H509-1976 were originauy ref erenced in the UFSAR. These standards address the physical properties and performance requirements for the control center adsorber filter charcoal. The performance requirements of the new standard are slightly lets stringent than the older 1976 stcndard. The differences are as follows:

1. The 1980 standard for methyt todine requires testing at 30 degrees C whereas the 1976 standard requires testing at 25 degrees C.

2. The 1980 standard for elemental iodine requires a retention accept 4nce criteria of 99.5% v;hereas the retention acceptance criteria of the 1976 standard is 99%.

For the methyl iodine test requiremen*s, increasing the test temperature to 30 degrees C t r,c re a se s tne removat efficiency of the carbon. However. 30 degrees matches the actual charcoat service conditions and le the industry standard approved by the NRC for testing control room emergency filtration afstems. The new charcoat performance standards are tqual to or better than the older requirements for the atomental lodine test requirements.

There is no change to plant eauipment or charcoat. The newly purchased replacement charcoal ess entially meets the same design criteria as the used charcoal but is tested to slightly different criteria. The use of the newer charcoat testing criteria will assure that the control center filtration system charcoat adsorbers will perform as designed. The new testing criteria supports the same service as the old testing did.

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SAFETY EVALUA710NS-PROCEDURES. TESTS AND EXPERIMENTS Page.31 SAFETY EVALUATION SUIAMRY Safety Evaluation No: 91-0066 UFSAR Revision No. 5 Ref erence Doctanent DER 91-0160 Soution(s) 3.3 Table (s) N/A __

Figure Change i 1 Yee IX) No Title of Change: Revision to UFSAR Tornado Loading Discussion SOMAARY:

This evaluation justifies rewording UFSAR subsection 3.3.2.3.7.2 to attow temporary non-seismic structures within 100 ft of t$e south watt of category 1 structures. This agrees with the current plant practice of locating temporary trailers and staging of construction materiets outside of the reactor buttding during outhges. Rewording UFSAR subsection 3.3.2.3.7.2 eliminates the need to write a saf6ty evatustion every time temporary structures or materiet is placed near the plant.

Category I structures have been designed to resist the impact forces of tornado-generated missiles pec Regulatory Guides 1,76 and 1.176. The Fermi 2 Safety Evaluation Report concludes that the plant design for externally generated missites is acceptable. The missile protection design does not take credit for not having may structures within 100 ft of the the south watt of Category 1 structures. The analysis in UFSAR section 3.5.4 Darrier Design Procedures, ervelopes the potential missiles that could be generated from temporary trailers -and other outage related materials. Engineering probabilistic analyses of tornado . missile hatards due to penetrations and openings in the reactor building and auxittery building watts, tornado missile camage to the residuat-heat removal complex cooling towers, and spent fuel poot tornado protection are not changed by the addition of temporary trailers and construction materlats outside of the reactor building.

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i SAFETY EVALUATIONS PROCEDURES, TESTS AND EXPERIMENTS Page 32 SAFETY EVALUATION SlANARY Safety Evaluation No: 91 0070 REV 1 UFSAR Revision No. 5 f<e f erence Document : LCR 91.126-UFS Section(s)

Tabh( s ) 0.3-2: 6.3-3 8.3-4 0.3+6: 8.3-7 Figure Change 1 J Yes IX] No 1

Title of Change UFSAR Emergency Dieset Generator (EDG) Lnad Table changes SlM4AW:

This evaluation justifies revising the UFSAR EDG toad tables to reflect changes in calculated EDO toads. The new calculation utilizos load data from other dogigncalculationeandapplicabledesigndocuments. In addition. Cable losses (1 R) were added to att loads and the swing but toads (automatio loads from MCC 72-CF) were added to EDO 14. As a retutt. the peak toads for the following scenarios changed as f ollows:

1. For a loss of offsite power (LOOP) at 0 to 10 minutes, the highest calculated EDO toad increased from 707 KW to 777 KW,

2. For a LOOP after 10 minutes, the highest calculated EDG toad increased from 2649 KW to 2705 KW.

3. For* a LOOP coincident with a toss of coolant accident (LOCA) at 0 to 10 minutes. the highest calculated EDO toad increased from 3030 KW to 3124 KW.

4. Foss a tOOP coincident with a LOCA after 10 minutes. the highest calculated EDO toad increased from 2846 KW to 2902 KW, The highest calculated toad. 3124 KW. occurs during a LOOP coincident with a LOCA at 0 to 10 minutes on EDO 14 when att EDOs are avaltable. This toad is within the shoet time rating (3135 KW) of the EDas and is, therefore. In compliance with paragraph C.2 of Regulatory Guide 1.9 revision 2 and item 347.2 of IEEE Standard 357-1977.

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SAFETY EVALUATION SLEAMRY Safety Evaluation No 91-0074 UFSAR Revision No. 5 Referenem Document: LCR 91 130-UFS Section(s) DA.4 Table (s) N/A Figure Change ( 1 Yes [XI No Title of Change: Residual Heat Removal (RHR) Complex and Condensate Storage Tank (CST) Yard Fire Analysis SUATMRY:

This evaluation justifies a revision to subsectione 9A 4.3.1 and 9A 4,7.2.1 of the UFSAR Fire Hazards Analysis to include analyses of a yard fire near the RHR complex and the CSTs. The previously approved fire hazards analysis did not address exterior hazards to safety related buildings and tanks. The RHR camplex analysis demonstrates that an oil spilt fire is the worst c.sse scenario for a yard fire and that it wiu not adversely affact the RHR complex. The CST analysis includes a discussion of the diked area around the CSTs and how it prevents an exposure fire in the yard from affecting the tanks.

Revising the Fire Hazards Analysis by documenting the effects of a postu'.ated fire in the yard adjacent to both the RHR complex and the CSTs is in accordance with the Generic Lettee B6-10 guidance on yard fire analysis. The original Appendix R analysis remains valid and unchanged. These discussions provida

• additional information to demanstrate the accuracy of the Appendix R yard fire assumptions. The RHR complex analysis concludes that a yard fire cannot spread to the 590' elevation of the RHR complex and. therefore, will not damage safe shutdown equipment or circuits in the RHR complex. The CST wnalysis concluwen that the CSTs are acequately protected from a potential exposure fire in the yard by the diked area surrounding them. In addition, the suppression pool can be used as an alternate water source if the CSTs are damaged in a fire.

SAFE 7Y EVALUATICXS PRDCEDURES, TESTS AND EXPERIMENTS Fage 34 SAFETY EVALUATION St24AARY Safety Evaluation Hot 91-0075 UFSAR Revision No. 5 Reference Document: LCR 91-135-UFS Section(s) N/A Table (s) 7.5-2 Figure Change [ ] Yes IX) No Title o' Change: Corrections to UFSAR Table 7.5-2 " Safety-Related and Power Generation Disp 1.ay Instrumentation" SUhNARY:

This evaluation justifies miscettaneous corrections to UFSAR Table 7.5-2,

" Safety-Related and Power Generation Display Instrumentation" to make it agree with other Fermi 2 base configuration design documents (DCDD). These changes include corrections to instrument numbers, design clasw OA levol/ seismic categories, number of channels, alarm setpoints, instrument ranges, anct instrument accuracies.

No physical modifications tc the plant were required as a result of these corrections. These corrections do not affect any DCDDs or the postulated accidents and accident analyses described in UFSAR Chapters 6 and 15. There la no reduction in the margin of safety as stated in the UFSAR. NRC Safety Evaluation Report, or Technical Specifications.

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SAFETY EVALUATIONS i PRCCEDURES. TESTS AND EXPERIMENTS Page 35 SAFETY EVALUATION SW4ARY Safety Evaluation Not 91-0082 UFSAR Revision No. 5 Reference Document: DC 4308 _

Section(s) 0.3 Table (s) N/A Figure W nge i 1 Yes (X1 No Title of Change: Delet ion o f 4160V Ground Faul,t Retsy Se t tings in UFSAR StM4ARY:

This evaluation justifies deleting the 4160V ground fault relay settings in UFSAR subsection 0.3.1.1.12.2. During a review of design calculations, an in.tonsistency between UFSAR subsection 8.3.1.1.12.2, " Circuit Protection", and tne actunt 4160V ground fault relay settic.gs was identified. The UFSAR stated that the relays are calibrated to operate with 0.5 amos secondary current, which is equivalent to 6 amps primary current . In practics, the relays are calibrated to operate at 15 ampe primary current . This change eliminates the conflict between the UFSAR and the field settings and removes extraneous and misleading information.

This change allows the use of a more realistic calibration setpoint higher than the value previously stated in the UFSAR. The ororation of equipment important to safety is enhanced because less equipment trips due to nuisance ground fault trips will be experienced. The field settings are still adequate to protect the power source and do not promote common mode fattures such as overcurrent

. induced fires or a mejor feutt.

SAFEVY EVALUAVIONS PPOCEDURES, TESTS AND EXPERIMENTS Page 36 SAFETY EVALUATION SIMAARY Safety Evaluation No: 91-0083 UFSAR Revision No. 5 _ _ .

Heference Document: LCR 91-149-HFS Sectlon(e) 8.2 Table (s) N/A Figure Change [ ] Yes [X] No Title of Change: 120 KV Dus Maximum Continuous Voltage Change SUhWARY:

This evaluation justifies reducing the 120 KV Bus maximum continuous voltage stated in UFSAR subsection 8.2.2.5.1 from 128 KV to 126 KV. This change is based on recent short- circuit studies identified by the Detroit Edison Transmission planning Department. These studies indicate that the maximum voltage may be reduced due to changes in the system load; network and capacitor changes that have made higher voltage excursions less possible; end improved' system voltage control.

This revision witt not change the available voltage on the downstream busses because the transformers will maintain bus vottage. The load flow and stability analyses previously evaluated in the UFSAR remain unchanged. The results of a design calculation indicate that att safety rotated busses are capable of performing their safety functions within design timits. Reducing the maximum overvoltage value reduces the potential f or equipment overvoltage and overcurrent. The new maximum voltage is within.the previously evaluated 120 KV/4.16n KV tennsformer SS C4 toad tap changer setpo11;1 and the reduction in the maximum source voltage does not have any impact on the 4160 V undervottage trip setpoints.

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SAFETY EVALUATIONS PROCEDURES, TESTS AND EXPERIMENTS Page 37 SAFETY EVALUATION SLAWY Safety Evaluation No: 91-0085 UFSAR Revision No. H/A Reference Document: NPP 20.107.01 Section(s) N/A Table (e) N/A Figure C Mnge ( ) Yes (X) No Title of Change Procedure Change to Allow Manual Initiation of Standby Feedwater (SBFW) System During Loss of Reactor Feedwater Pump (RFP) Reactor Recirculation Pump (RRP) Runback Concurrent With Loss of Heater Drains

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This evaluation justifies the revision of abnormal operating procedure (AOP) 20.107.01 " Loss of Feedwater or Feedwater Controt", to attow manual initiation of the SSFW system during a RRP runback caused by the loss of one RFP, a coincident loss of heater drains, and reactor power greater than 70%. An evaluation of acceptance test data for the design change t ha t allows the Fermi 2 to be operated in the maximum extended load line limit (MELLL) region indicates-that when the reactor is operating at 100% power and.75% ce?e flow (the upper end of the MELLL region) a RRP runback caused by the loss of one RFP and heater drains will run back power to 73%. Since the maximum faedwater flow of a single operating RFP can only maintain 70% reactor power, the steam /feedwater flow mismatch will lead to a level 3 scram in appro/imately 6 minutes.

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In order.to avoid a scram. operating procedure ACP 20.107.01 has been revised to direct the operators to:

1. Use SBFW to avoid a reactor pressure vesset low water levet scram on a loss of one RFP. a coincident loss of heater drains, and reactor power greater than 70%.

2. Insert the cram array controt rods to the extent necessary to reduce reactor power to 68%.

3. Terminate S8FW on any inadvertent high pressure coolant injection (HPC1) initiation.

SAFETY EVALUAT10NS PROCEDURES. TESTS AND EXPERIMENTS Page 38 Safety Evaluation No. 91-0085 (continued):

The use of SBFW reduces the probability of a reactor scram that would challenge equipment important to safety. The injection of SBFW at 1200 gpm is a cold water addition which rebutts in a positive reactivity insertion. However, the margin of safety is not reduced because this reactivity addition is bounded by the 5000 gpm cold water (40 degrees F) HPCI reactivity addition analysis. The procedural step to terminate SBFW if HPCI is inadvertently initiated ensures the reactor will be operated within the HPCI analysis. If reactor core injection cooling (RCIC) is inadvertently initiated white SBFW is in operation, the HPCI addition analysis is still the bounding analysis since RCIC will only inject an additional 600 gpm. The limiting transients fwr a reactivity insertion are the turbine generator trip at 100% power and the feedwater centro 11er failure at power levels betew 100%. Therefore, the use of S8FW as described in this procedure change is not a limiting transient and has no advceso impact on safety.

SAFEVY EVALUATIONS PROCEDURES. TESTS AND EXPERIMENTS Page 39 SAFETY EVALUATION SLM MRY Safety Evaluation No 91-D086 UFSAR Revision No. 5 Reference Docu#eent: LCR 91-152-UFS Section(s) 9.5 Table (e) N/A Figure Change ( 1 Yes IX) No Title of Change Revision of Eme'gency Dieset Generator Fuet 011 System operation in UFSAR Subsection 9.5.4.2 SLM4ARY-This evaluation justifies revising UFSAR subsection g.5.4.2. The following changes have been made

1) The description of the startup of the alternate fuel oil transfer pump and alarm has been changed to indicate that they do no t initiate at the same level. Tha transfer pump start signat occurs at a higher levet than the alarm.

2) The statement that the EDG-fuet oil transfer system strainers are changed out has been changed to state that the filters are blown oown.

The strainers are Leslie self-cleaning "Y" strainers. When the strainer blowdown valve is opened, the strainers ese designed to allow fuet oil to flush out the sediment collected on the-screen.

-3) The time available to take corrective actions has been changed from the running time of a futt day tank to the running time of the fuet inventory at the low alarm levet in the day tank. This reflects the time available to take action for the worst case scenerio; both f strainers plugged and the day tank low levet alarm in.

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This revision does not change the design, function, ce operation of the EDos.

l Long term sustained operation of the EDGs is maintained. The ott used to blow down the EDO fuel oil transfer system strainers cannot be used as fust for the EDGs. However. the amount of olt used is insignificent (1 to 2 gettons per j cleaning) compared to the 35.280 ganon fust supply providect to maintain operation f or - seven days. Therefore, bl.ow ing down the strainers does not l-affect the abnity of the EDGs to run for the required time intervat.

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SAFETY EVALUATIONS PROCEDURES, TESTS AND EXPERIMENTS Page 40 MFETY LVALUATION !NMRY Safety Evaluation H2: U1 0089 UFSAR Revision No. 5 _

Ref erence Docurnent: LCR 91-158-Ur3 Section(s) 8.2 Table (s) N/A Figure Change i 1 Yes IX) No Title of Changer Removal of the Minimum Nazimum Dus Voltage Limit Table in UFSAR Subsection 9.2.2.5.1 and Grid Configuration Study Year Revision S1MAARY:

This evaluation justifies removing the minimum /memimum bus voltage limit table from UFSAR subsection B.2.2.h.1 and revising the year that the latest grid configuration study was performed. The minimum / maximum bus voltage limit table was removed for the following reasons:

1. Calculations show that each bus has a different type of loading, total loading, and feeder length. Inerefore, different vol. t a g e limits apply.

2. Electrical loading has varied over the years and the voltage limits set during construction are not the same during plant operation.

Fixed voltage limits for all buses are no longer relevant to the operating plant environment.

3. Voltage limit evaluation in an ongoing process.

4. Bus voltage limits are determined and controlled by design calculations. Design calculatione evaluate plant bus volt ages which allow proper operation of all safety related electricat equipment and plant process systaan. Design calculations consider the minimum AC voltage and current pickup values at the 120 V levels.

The year of the gr configuration study has been changed from 1981 to 1991 to reflect the updated tudy.

The removal of the voltage limits table does not change relay types or settings. Pomoving the voltage limits table from the UFSAR has no technical impact on the operation of the equipment since existing calculatione control the equipment voltage 1,1mi t s . All safety related loads remain capable of performing their safety functions. The analyses, equipment, instrumentation, and voltage limits identified in the Technt:a1 Specifications are not changed.

The identification of and response to a degraded grid condition described in UFSAR subsections 8.2.2.5.2 and 8.2.2.5.3 are not affected by this revision.

SMETY EVALUATIONS PROCEDURES, TESTS AND EXPERIMENTS Page 41 SAFCTY EVAL 11AT10N SUW WHY Safety Evaluation No 91-0000 UFSAR Hevision No. N/A Heterence Documents NPP-CRIIPP-91 011 Section(s) NL

-012.and .013 __ _,

Tabte(s) N/A ___ _ _ _

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Figure Change I 1 Yes (X) No Title of Ctangus Control Rod Stade Processing SLMAARY:

This ovatustion justifies the activities requtrod for procesetng the controt rod blades.(CRD) removed from the reactor during the second refueling outage.

'This safety evaluation covers (1) activities prior to and following processing actions such as toad handting and rigging; (2) CR0 processing within the spent fust poot - (SFP): and (3) the effects of the activities on SFP cooling. The CRas aero processed one at a time. The CR0 stettite bearings and velocity timiters were sheared off using spent fust poot (SFP) curb mounted equiswnent .

After being compacted, each CR0 was inverted and placed inside a liner. The fitted 11nero are stored in the spent fust poot.

Crane hoists were selected with adequate capact.'.y and double the norme't safety factors to provide compliance with regulatory requirements. Safe load paths were used to move the Crus-and associated process equipment within the SFP area. With-the exception of the CRBs. no equipment was moved over the spent fust encks. The offeats of a CRB deop over the opent fust racks la bounded by the SFP fust assembly drop analysia. The impact energy is below the UFSAR enalysis impact energy of 2000 f t-tbo because the CRBs are approntmately ,

one-third the weicht of a fust assembly and the minimum water depth maintained-over the CRDs'is limited by procedure NPP-CRDPP-012 to ensure that the CRS drop height is tems than the fuel assembly drop height. Similarly, the effect of a ,

CRD drop on the SFP Liner is bounded because the fuel bundle is henvise than a CR8 and the drop heights are similar. The components related to CR0 processing are seismicatty qualified and the tonding effects-of thess components on the SFP supporting structures is acceptable. The impact of CRB processing does not have any significant effect on SFP cooling.

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hAFETY EVALUATIONS PROCEDURF.0, TESTS AND EXPER!MENTS Page 42 l SAFETY EVALUATION SULTAARY Safety Evaluation No: 91-0091 UFSAR Revision No. 6 Reference Document: DER 87-398 Sootton(e) 6.4 Table (s) N/A I

Figure Change ( 1 Yes IX) No ._

Title of Change: UFSAR Clarification for Control Center Heating, Ventitation, and Air Conditioning (CCHVAC) i SLANVtY:

This evaluation justifies revising UFSAR subsection 6.4.2.3.1 to clarify the fact that the CCHVAC emergency intake isolation dampere remain closed when chlorine gas is detected. The original wording stated that these dampers "would close' when chtorine gas is detected. This wording incorrectly implied that the emergency intake isolation dampers can undergo a lineur change when chlorine is detected, The emergency intake isolation danpers are required to stay closed during chl.orine release accidents if .he CCHVAC system is in the normat, purge, we chlorine mode at the onset of the release, if the CCHVAC system is in the recirculation mode (emergency air intake dampers open) closure of the dampers is not required to koop chlorine concentrations down to En acceptable lavet in the controt room, Therefore, there is no omergency air intake damper closing operation associated with the detection of chlorine gas ,

and the resultant alignment of the CCHVAC system. This UFSAR revision is the result of findings in Technical Specification Improvement Program item N749.

This UFSAR revision clarifies the operation of the CCHVAC sys t ern. This revision does not change its design, function, or operation. The CCHVAC system still conforms to the Regulatory Guide 1.95 requirement that the controt room chlorine concentration shoul,d not exceed 15 ppm within two minutes efter the operators are made aware of the presence of chlorine.

SAFETY EVALUATIONS PROCEDURES. TESTS AND EXPERIMEN19 Page 43 SAFETY EVALUATION SUW MRY Safety Evaluation No 91-0003 UfSAR Rovision No, b Reference Document LCR 90-17P-UFO Section(s) 7.3 __

Table (s) N/A _

Figure Change ( 1 Yes (X) No Title of Changet Removat of Reactor Water Lovet 1 Isolation Levet in UFSAR Subsection 7.3.2.2.7.1 StA W JtY:

This evaluation justiften esmoving the reactor water 1.evet 1 isolation level, setpoint in UFBAR subsection 7.3.2.2.7.1. This subsection contained a sentence stating that the reactor water levet 1 isolation level is approximately 14 inches above the top of active fuet. The current Tecnnicet Specifications (Table 3.3.2-2 afd 81gure B 3/4.3-1) bhow a trip setpoint of greater than or equal to 31.8 inphes and an allowable value of greater than ce equal to 24.8 inches. A design calculation 'and Cenorat Electric Specification 22A2019A8 agree with the Technicat Specif tentions. The origin of the 14 inch setpoint.

and whethee it is raeant to espresent the nominst trip setpoint, a nowable value, or the. analytical Limit is uncienr. Removal, of this setpoint from UFSAR subsectien 7.3,2,2,7.1 provides consistency as the values for the reactor water levats 2 und 3 are ,.ot'found.in the tout.

This revision coes not have any effect on-the design, function, of operation of the plant. The autual values for the reactor water levat 1 isolation .tevet remain unchanged in the Technical Specificatione and design catouto*lon.

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SAFETV EVALUATIONS PROCEDURES, 1ESTS ANO EXPERIMENTS ]

Page .54 l MFETY EVALUATION SLA&%RY Safety Evaluation Not 91-0094 UFMR Revi sion Na, 5 Reference Document: LCR 91 -165-U3 Section(s) 13.4 Tatde(s) N/A __

Figure Change [ } Yes (X) No Title of Changet UFSAR Chapter 13 Revisions SLEMNtYt This change (1) revises UFSAR subsection 13.4.3.2 to remove the requirement that the secretary of the Nuclear Safety Review Group (NSRQ) be appointed from the membership of the NSRQ; and (2) revises UF5AR subsection 13.4.3.3 to remove the requirement for the Independent Safety Engineering Group (ISEO) to make detail,ed reconnendationa to the chairman of the NSRG.

Eliminating the requirement that the NSRQ secretary shall be a member of the NSRO is administrative in nature. There is no change to any system, structure, ce component. No now mode of plant operation is introducet NSRG staffing requirement s 'and att other requirements of Technical Specification 6.5.2 are not changed by this revision.

Eliminating the requiremen. that the IDEQ w,ake detailed recommendations to the chairman - of the NSRG makes UFSAR subsection 13.4.3.3 consistent with the requiremsnts of Technical Specification 6.2.3, As amended by NRC approved License Amendment.63, this technical specification only requires that ISEO make recormnendations to the Vice President . Nuclear Engineering and Serv!,ces. The ISEQ staffing and reporting requirements are not altered by this change.

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UFSAR Revision No. 5 Ref erence Doctament s LCR 91-024-UFS Section(s) 7.73 9.43 10.43 11.3 Tabte(s) 11.3-1: 11.3-2; 11.3-3.-

11.3-4: 11.3-5: 11.3-1 Figure Change [ ] Yes [XI No Title of Change: Incorporation of Operation of the Off-Gas System at Flowcates Greater Than 40 scfm into the UFSAR StnHUW This evaluation justifies incorporation cf off-gas system operation at flowrotes greater than 40 s:fm into the UF*LAR. Per deviation event report,

' DER 91-0598, an engineering - functionat analysis was performed to support the long term operation of the off-gas system with condenser air intenkage flow in excess of the 40 scfm UFSAR value. This analysis concluded that air flows greater than or equal to 80 sefm have no adverse impact on the functioning and performance of the equipment, associated instrumentation, and accessories. The

. analysis further concludes that the limiting factor for the capacity of .the off-gas system is the capacity of the ring water vacuum pumps when operating both pumps in parattet. As a result, UFSAR sections 9.4, 10.4 and 11,3: and-tables 11.3-1, 11.?-2, 11.3-3, 11.3-4 and 11.3-5 have been revised to state that: -(1) The off-gas system is capable of processing off-gas with air flows greater than 40 scfm and, (2) The off-gas parameters will vary in the event of air.ftows greater than 40 scfm.

In addition, various inconsistencies and redundancies were removed f rom UFSAR sections associated with operation and analysis of the off-gas system. As a result, the following changes have been made:

1. UFSAR subsection 11.3.2.7.5 has been revised to_ change the menon residence time in the charcoal adsorbers from 14 to'16 days. The derived residence time for menon in the charcoal adsorber is 17.3 to 18.9 days. Therefore. it is.scceptable to conclude that the residence time design basis value is 16 days.

2. UFSAR subsection 11.3.3.4 has been revised to change the offsite dose limit from 0.17 rem to 0.5 rom. The original Safety Guide 26 specified an offsite dose limit of 0.17 rem. However, the UFSAR commits Fermi 2 to the 0.6 rem limit of Regulatory Guide 1.26.

SAFETY EVALUATIONS PROCEDURES, TESTS AND EXPERIMENTS Page 46 Safety Evaluation No. 91-0090 (continued)1

3. UFSAR subsections 11.3.3.4 and 11.3 3.5 have been revised to delete the offsite doves that result from a total failure of the offgas system. The offges system failure analysis is provided in UFSAR section 15.11.

4. UFSAR subsection 11.3.2.7.3.1 has been revised to state *5at the value of the dynamic adsorption coefficient, K , and the raidence time are determined emperimentally or derived per calculation. The UFSAR previously stated that K and the residence time were only determined experimentally.

5. The chiller outlet off-ges temperature in UFSAR Table 11.3-4 has been changed from less than or equal to 4 degrees F to 14 degrees F. The former number did not agree with the numbers stated in 'J F SAR subsections 11.3.2.7.5 and 11.3.3.3.8, operating experience, or charcoal, adsorber tes t s.

6. The ranges of charc oat filter outlet flow transmitter N530 and off-gas charcoal units to edsorber filters pressure transmitter N525 in UFSAR Table 11.3-5 have been changed to reflect the current ranges in PDC 8471, revision C. and the central, component data base.

7. UFSAR subsection has been revised to change the off-gas high flow alarm from 50 sofm to 55 to 70 scfm. This setpoint was revised per engineering design package, EDP 11816.

8. Procooler and chiller outlet temperatures have been deleted from UFSAR subsections 11.3.2.7.5 and 11.3.3.3,6 as this information is already in UFSAR Table 11.3-4 In addition, UFEAR section 11.3 has been revised to provide clarification and reference for the above changes.

The operation of the off-gas system with an air flow greater than 80 scfm has no adverse-impact on the function or performance of the equipment, associated instrumentation, or accessories. The radioisotope inver' tory values that ssult from this change are still within the existing bounds in UFSAR section 11.3 and the accident analysis in section 15.11. Offsite doses will remain below NRC limits and will also be ALARA. Adequate safeguards and procedures exist for detecting and limiting of f site doses well in advance of approaching any dose limits as the Fermi 2 Offsite Doae Celculation Manual provides the methods for both release rates and offsite dose analysis; technical specifications and administrative controls limit the off-gas release rate. The Failed Fuel Action P1,an provides the required ALARA safeguards.

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SAFETY EVA1.UAT10NS PROCEDJRES TESTS AND EXPERIMENf3 Page 47 SAFL7Y EVALUAT10H DULt.MRY Safety Evaluation Not 91 -C' 107 UrtAR Revision No. $_

Ref erence Doctsnent LCR 91-190-UFS Coction(s) 9.1 Table (e) N/A 1

Fipure Change I 1 Yes (X) No Tide of Change Discussion of Fermi 2 Specist Lifting Devices to the UFSAR t

SumMHY:

This evaluation justifies adding a discussion of special lif ting devices and their periodic testing in UFSAR subsection 9.1.4.4 A 1pectat lifting device is a device designed specificetty for handling a certain type of load. The Formt 2 spectat lifting devices are the reactor presevre vesset head strongback and the dryer / separator lifting device. Previously, the liFSAR did not provide e discussion of specist liftirg devices or any guldence on their periodic testing. This change reelses UFSAR subsection 9.1.4.4 to state that (1) the

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Fermi 2 spectat liftino devices meet the criteria of NUREG.0612, " Control of Heavy Loads at Nucteer Power plants" and, (2) Fe rrni 2 meets the tonting guidelines of NUREG-0612 by performing the load bearing weld futt non-destructive examination option of ANSI N14.6 1978 at five year intervata.

A visual inspection of these devices is performed each your and before each peelod of use.

The addition of the discussion of special lifting devices in the UFSAR does not affact existing plant equipment or change operating procedures. No new equipment, modifications, or testing is introducert by this chance. The special lifting devices and their testing ere in conformance with NRC guidelines in NUREG-0612.

SAF ETY IYALVATIONS PROCEDURES, TESTS AND EXPERIMENTS Page 48 EAFl'lY EVALUAT10N MMMRY Safety Evaluation No 91-010,9 Uf!AR Revision No. 6 Ref erence Doctament: LCR 91-193-UFS Section(s) 9.1 Tabte(s) N/A ,

figure Change ( ) Yes (X) No Title of Changet High Density Spent Fuel Racks Neutron Absoaber Materiet Survoittance Program UFRAR Referenr.e Change SG MARY:

This evaluation justiftee changing the UFSAR reference for the high density spent f ust racks Doref ten neutron absorber materiet survoittance program from TM-686, " Joseph Ost Corporation, Licensing Input on High Dentity spent Fuet Rocks for Fermi 2 Project" to Northoest Technology Corporation (NETCO),

"Revloed Doraftex Coupon Surveittance Program Document No. 073-01". This survettlance program monitore the condition of the Boreflex neutron absorbing materlat in the spent fust storage rocks. The new reference describes how the neutron absorber materlat survaitlance program in currently carried out. The Joseph Oat Corporation program cetted for periodic measurements of the Doraflex coupone dimensions to be taken over the life of the plant and compared to besetine data. The baseline cate was to be taken by the Joseph Oat Corporation during fabrication and prior to coupon irradiation. However, the baseline data has been taet. As a reeutt, a DER was written to document het the Joseph Ost Corporation program could not be property implemented without the basetino data. The DER corrective action contracted NETCO to take current dimensions of the Borafb x coupons. Future measurements will, be compared to this new besettne data. A new in-service testing program for testing the Doreften coupons by NETCO was developed to replace the Joseph Ost Corporation program and is now referenced in UFSAR subsection 9.1.2.2.2.

The NETCO program makes the following additional changes:

1. The method of fastening the coupon housing f ror,t , conter, and back plates has been changed. Formerty. the plates were tack welded together. The front cover plate currently has deformebte capture tabs that are bent around the back of the conter plate. The conter plate is still tack welded to the back plata.

2. Two of the three untrradiated coupons that were removed prior to the first refueling outage will be reinstatted in the spent fuel pool.

These coupons will be removed prave to the fourth refueling outage and tested. These coupons witt aid in determining the maximum anticipated Boraften shrinkege in the high density fuel storage racks.

I

SAFETY EVALUATIONS PROCEDURIS, TESTS AND EXPERIMENTS Page 49 Safety Evaluation No. 91-0109 (continued):

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3. Th9 three irradiated coupons removed prior to the second refueling nutage win be reinstelbd in the spent f ust pool. These coupons will also aid in determining the maximum anticipated Boreften shrinkage in the high density fust storage rocks.

4. The NETCO testing progrum utitires a radionssey of the coupon surfacu for beta and gamma radiation. This provides an indication of the extent of water permention in the Boreften.

5. The NETCO program does not require neutron radiography of the coupons as this is not a quantitative measure end there are smatt variations in the boron-10 tonding in Boraftem.

G. The NETCO program changes the number of coupone removed per testing int erval f rom three to two. This attows the Boraf'.ex coupons to be removed throughout the design service life ci the fust racks as opposed to the eight year period under the Joseph Ost program.

These changes are limited to the high density fust storage rack neutron absorbing materlat surveillance program. These changes restore the program to a functional status and meet the program's design requirements. The NETCO survel u ence program follows the EPRI guidelines for Boraflex surveittance programs. The weight of the survoittance specamen tree which holds the sample coupons remains weit within the bounds of the f uel handling accident analysed in the UFSAR. These changes do not impact the seismic qualifications of the fuel rocks. Technical Specification 6.8.5.d requires that a program be established. implemented, and maintained to assure that any unanticipated degradation of the high density spent f ust racks wil,1 be detected and will not compromise the integrity of the racks. The requirements of this technical specification are not impacted by those changes.

SAFETY EVALUATIONS PROCEDURES, TESTS AND EXPERIMENTS Page 50 i

SAFETY LVAlt1ATION Sl8AMRY l

Safety Evaluation No: 91-0117 UFSAR Hovision No. 6 Reference Document: LCR 91-196-OfS Section(s) BA.6 ,

Tabte(s) N/A _

l Figure Change ( ) Yes .X) No Title of Change: Revisions to UFSAR Section 9A.6, "Pire Protection Conditions for Operations" i SmeA8J1Y:

This evaluation justifies revising UFSAR Sect.on 9A.G. " Fire Protection Conditions for Operation" to bring fire protection systems surve il.tanc e intervate into closer confoemance with National Fire Protection Association (NFPA) codes: facia' ate the implementation of compensatory measures and correct typogenphicat errors. The changes are as f ollows:

1. The fire detection instrumentatiori funesional test turveillance intervat in UFSAR subsections 9A.6.1.2.1 and 9A.6.1.2.2 have been changed from six months to twolve months. The twelve month interval meets the functionat testing intervat requirements of the 1990 edition of NFPA 72E.

2. UFSAR subsection 9 A . 6.1. 2. 3 ha s been deleted. This section covered non-supervised fire detection alarms circuits. It was deleted because there are no non-supervised fire detection storm circulta at Fermi 2.

3. Typographical errors in UFSAR subsection 6A,6.3.1 have been corrected. These changes are editorial and do not affeet the Fire Protection Program.

4 A note has been added to UFSAR subsections 9A,6.3.1 (spray and sprinkler systems), 9A.6.4.1 (carbon dioutda systems), anst 9A.6.5.1 (haton systems) to indicate which of these fire supp'seulon systems are in fire zones containing redundant safe shutdown equirrnent. This change attows easier and more reliebte identification of aroes that contain redundant safe shutdown equipment and also ensures that the proper fire watch (continuous or hourty) will, be assigned when a fire suppression system is declared inoperable.

SAFETY EVALUATIONS PROCEDURES, TEbTS AND EXPE91W NTS Page 51 Safety Evolustion No. 91-0112 (continued):

5. The menuet cable spreading room sprinkler system has treen removed f rom UFSAR subsec tion 9A.6. 3.1. This room alread/ has a halon suppreeston system and there is no requirement to have backup manual suppreselon tystems to protect plant systems or equiseont.

6. The puff test requirement for UFS.AR subsections WA.6.4.2.2 (carbon dioulde systems) and 9A.6.8.2.1 (haton systems) has been deleted.

This test has been deleted becerse it does not demonstrate system operability and is not required by the NFpA 12 and 12A.

This revision does not impact the operability or function of any fire protection component. The changes do not 6ffect the operation, function, or rettability of any plant system or component. The consequences of an accidental, release of a fire suppression agent (watee. CO ' r h"\D")

2 reeutting from the fatture or inadsortant operation of a fire suppression system is anatyred in UFSAR subsection 9.5.1. The conseq;ances of fire protection system or component matfunction in addretted in UFSAR section 9A.6.

These changes do not alter the 1.evet of protection provided for the plant systems addressed in the Formi 2 Technicet Specifications or Benes.

SAFETY EVALUATIONS PROCEDURES. TESTS AND EXPERIMfHTS Page 52 SAF ETY EVALUAi10N StMJARY Safety Evaluation Hot 91-0116 UfSAR Revision No. 6 Re f erence Document t (CR 91-197-008 Section(s) 8.31 9.4 Table (s) 9.4-81__9.4-11 l

Figure Change ( 1 Yes (X) No Title of Changet Miscenaneous Co< er*1an -

% '!:ations tc the UFSAR Discussion of it a Re ad 4' Ef L ; ..smovat (RHR) Complex Ventitation System i

M&tWVlY This evaluation justifies enaking corrections and clarifications to JFSAR sections 8.3 and 9.4 to resolve the findings of the Fermi 2 Independent Safety Engineering Group (ISEt4) Report 90-011. These changes are as foMows:

i

1. The design ambient temperature for the emergency dietet generator (EDG) rooms has troon changed from 125 F to 1*2 F in UFSAR subsection 8.3.1.1.8. UFSAR subsection 9. 4. 7.1 Safety Evaluation Repcet section 9.4.3, and a design calculation correctly state that the EDO room has a 122 F maximum ambient design temperature.

2. The reference to EDG switchgear room ventilation system local manual control switches in UFSAR subsection 9.4.7.2.5 has been removed because these switches do not exist.

3. The reference to control room indication for EDO room temperature and high HVAC fitter differential pressuto in UFSAR subsections 9.4.7.2.5 and 9.4.7.3.5 has been removed because this indication does not exist.

4. UFSAA subsection 9.4.7.3.5 has been revised to correctly state that the system logic automatically starts the pump room ventitation system on high temperature or EDGs running. The original text incorrectly stated that it required both the high temperature and EDGs running to start tre ventitation system.

5. The inappropriate references to the switchgear room under the row describing the pump room ventilation system in UFSAR Table 9.4-0 have been replaced with the words ' pump room", The switchgear room ventilation system is described elsewhere in this table.

SAFETY EVAL.UATIONS PROCEDURES, TE$TS AND EXPERIMENTS page 63 Saf ety 1:valu'ation No. 91-011B (continued):

6. The UrSAR Tabte 9.4 8 commente column for the purt'p room has been revised to state that the operator will teko the necessary actione depending on room t erope r a t ur e . This table previously stated that the operator will take the necessory actione for a fitter high pretours dif f erential switch fatture, However, there le no high differentist pressure indication in the control room.

7. The pump room ventitation system fan capacity has been changed from 7500 sofm to 12500 onfm in UFGAR Table 9.4-11. The tevloed capacity 14 1:ssed on vendor drawings and the plant component database.

These revisione do not change the components or the design requirerronte of the ,

RHR compteu HVAC system. These changes only clarify and correct the UFSAR descriptions of the RHR comp 1.en HVAC sysisen,

SAFETY EVALUATIONS PROCEDURES. 1E019 AND EXPER1MENTS Page 54 fW C1Y LVAlliAT10H N#.WHY Safety Evaluation Not 91-0119_ UFr.AR Movielon No. 6 Reference Occument: LCR 91 138-UFS Cootlon(e1 11.4 _

Table (s) 11.4-b; 11.4-6 Figure Change ( 1 Yes (X) Ho Title of Chege Removat of References to Gaseous Particulates and lodine in the UF3AR; Additionat Impeovementa to tbe Chemietry Program Descritted in the UFSAR k SLSAWiY :

The evaluation justifies removing references to gessous particulates and iodine end miscellaneous clarifications in order to strengthen the post accident samp\ing program (PASS) and improve the accuraev uf the UFSAR as it pertains to the Fermi 2 chemistry program. Formt 2 procedures do not require the use of gaseous particutete and iodine samples for attuanete measurement of containment radiation or for estimation of core damage. The misceULaneous changes are intended to reference state of the art instrumentation and procedures while providing geester flamibility to utilire suisting equivatont equipment. These changes also clarify the UFSAR to reflect teve system capabitaties and provide specific details as to the implementation of regulatory commitments.

Data gathered from a con t airpen t atmosphere particulates and lodine samole provides no input to the damage assessment calculations of Fermi 2 procedure

?6.000.15 " Determination of Extent of Core Damage". Therefore, the otiminetton of the capabit,ity to gather gaseous particulate and iodine samples does not degrade the ability to mitigate core damage. In addition, this capability to not identified in emergency plan procedure EP - 46, " Calculation of Estimated Containnent High Range Radiation Monitor SQTS/AY. Monitor Reevings if Ins t rurien t s are inoperable or Of f sento". Therefore, that change does not impact the Emsegency P\an. The miscettaneous changes do not create any physical or proceducat changes to the PAS $ system. 1hese changes are consistent with the comitment s identified in the NRC Fermi 2 SER Chapter 22.

section !!.B.3 and its subsequent supplements. These changes do not chettonge the conclusions of the instrument line break ennlysis in the UFSAR. These changse do not effect the training; sampling and analysis; or maintenance requirements of TecNiicet Specification 6.8.$.c. " Post. Accident Sampling".

SAFETY EVALUATIONS PROCEDURES. TESTS AND EXPERIMENTS Page 66 tiAFETY EVALUATION !.l8 MARY Safety Evaluation hos 92-0002 UFSAR Revision No. 5 Heference Document: LCR 93 11641FS Sectlon(e) 8.3 DA.E Tabte(s) N/A Figure Change ( ) Yes (X) No Title of Change: Revise the UFSAR to List the Exceptions to the Construction and Qualification Pequiremente for Special Wires and Cables SLMAMW This evaluation justifies revising the UFSAR to list exceptions to the construction and qualification requirements of specification 3071-0B0, "Special Wires and Cables". The fottowing onceptions have been addedt

1. Balance of plant (DOP) medium voltage cables utilized in the underground power supply

2. Wiring for the tighting, communications, and security systems

3. Internal controt panel wiring in the controt conter 4 Vendor supplied wiring The BOP medium voltage cables utiliked in the underground power supply include some 5 kv and 15 kv rated cables that are used to interconnect transf ormers and load centers that are outside of the main power block. These cables were obtained from Detroit Edison stock and cable cualification documentation does not exist.

The lighting. Communications, and security systems cabling it routed tnroughout the plant in separate enclosed raceways and generatty doea c't enter the cable tray systems. However, some communications cables are routed in BOP cable trays. They use flexible conduit for separation from the other cables. The tighting cables are Nationet Electrical Code NFPA type THHN or XHHW. Both insulation types are flame retardant systems. The communications cables are spectatty cables specified by the manufacturer. The security system cables are a similae in design to those that meet the 3071-080 specifications but their quellfications are undocumented.

SAFETY EVALUATIONS PROCEDURES. TESTS AND EXPERIMLNTS Page ti6 Safety Evaluation No. 92-0002 (continued):

The control conter internet controt panel, raring was purchased under specification 3071-080 but was not required to meet the radiation exposure and post accident (LOCA) environmentat quotification requiremente of IEEE-323.

However, they do meet the requiremente of IEEE-303. The insulation on these wires is much thinner than the wire ineutetton cetted for in specification 3071-080. However, after an accident, the centrol conter is a milder environment and, therefore, the thinner insulation win not adversely affect the associated systems.

4 The qualification of the equipment internet wiring was the responsibi\ity of the vendor ce the vendor's suppliers. Detroit Edison was responsible for reviewing the vendor's qualification documentation. There were no requirements to meet the safety re\sted equipment standards of IEEE 303 and IEEE 323 placed on the non-safety related equipment puachase specifications. The wire used was appropriate for the application. Equipment maintenance and modifications utiltre wire originsMy used by the equipment vendor or wiring that meets the 3071-080 specifications.

AM Fermi 2 cables are either certified to the flame testing requirements of IEEE 383-1974 or are fut\y enclosed in metallic raceway / enclosures with fire stopo es required by the 3071-080 specifications. The instaMation techniques used at Fermi 2 do not impWet the fire protection program. Att safety related cables are fully quellfied fe; ineir functions and for the areae in which they are tocated.

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SAPETY EVALUAT IONS PROCEDURES TESTS AND EXPERIMENTS Page 67 FAFETY EVALUATION StMAARY Safety Evaluation No 97-0009 __ Uf'SAR lievision No. $ _

Reference Docuenent LCR 92-018-UFS Sectiunta) N/A Table (s) N/A Figure Change (X) Yes [ ] No Title of Change Reautor Water Cleanup System (RWCU) Recirculation Pump B Motor Replacement SLA4AARY:

This modification replaces the motor for RWCU recirculation pump D. The motor was rep \ aced due to emcessive vibration. The new motor is rated at the same horsepower and voltage as the old motor ( 60 hp and 460 vac). The futt load current of the new motoe is 67 emperes whereas the original motor futt toad current is 60. 6 ampere s . Therefore, UFSAR figure 8.6 5 has been revised to change the fun load current on bus 72E POS. 2D from 00.6 amperes to $7 amperes.

RWCU recirculation pump B is in the non Q. seismic 11/1 portion of the system and does not serve a safety rateted function. No circuit modifications were required to accommodate the change. The aman dit f orence in fun load current between the original motor and the replacement motor has been evaluated.

Engineering considerations such as emergency dieset generator toeding, short circuit current, power feeder rating, protective relay setting, and power uprate impact, were reviewed and found acceptable.

umme -

faFETY EVALUATIONS PROCEDURES. TESTS AND EXPER1MENTS Page 58 l

i GAF ETY EVALUATION StMMMY ,

Safety I' valuation No: 92-0010 UFSAR Revision No. 5 /

Reference Documents _LCR 90-0?9-UFS __

Section(s) 9.2 Table (s) Nf_A Figure Change ( 1 Yes 1X1 No Title of Change: Deletion of Lined Piping Requirements in the UFSAR tlA4MRY:

This evetuation justifies deleting the material specifications for make-up dominereltger lined pipe and sampling lines M UFSAR Subsection 9.2.3.3. This subsection originalty stated that vessets and valves in contact with caustic or acid solutions are rubber lined and piping handling non-neutral flow is polypropylene lined. It atto stated that sample lines are Type 304 staintees steel. In the original make-up domineralizer system design att of the non-neutral f\ow piping was not polypropylene lined and att velves were not rubber lined. Some piping was staintese steet or heavy watted carbon eteet and some valves were polypropylene lined. As a result, the s t a t emerit s specifying rubber or polypaopylene lined components and Type 304 stainless stoet sample lines have been deleted. These statements are not necessary because atternate w,storials are available to handle non-neutral fluids. This change allows flexibility in choosing materlats compatible with the fluide based on standard engineering practicos.

This revision does not iritpac t the function of the make-up domineralizer system. Removing these specifications has no itapact upon the existing acoident scenarios contained in UFSAR chapters 6 and 16. The make-up domineralizer system does not initiate these events. The make-up domineralizer system is not required foe safe shutdown of the plant and no safety related equipment is located within close proximity of any of the make-up domineratirer componenta in question.

SAFETY EVALUATIONS PROCEDURES, TESTS AND EXPERjMENTS Page 69 SAFETY EVALUATION SLMARY Safety Evaluation No 92 0014 Urr.AR Revision No. 1.

Reference Document: LCR 92-019-Ur8 Section(s) 3.1: 4.5 7.1 7.6 Toble(s) N/A Figure Change tXI Yee [ ] No Title of Change Removal of Neutron Startup Sources St2 MARY:

This evetuation justifies revising the UFSAR to remove statements that startup neutron sources are installed in the reactor core. The startup neutron sources were removed after the first refue\ing outage. They are not required in a sufficient \y irradiated core and Generat Electric Service Information Letter SIL No. 215 recommended that the startup source holders be removed. General E\ ectr.o based its recommendations on the fact that broken startup source holders had been discovered in seven DWRs. The earliest feiture was observed after two fust cycles.

The startup neutron sources do not impact the minimum neutron count rate required by the source range monitors for proper startup operation. The function of the tiertup neutron sources has been replaced by the irradiated fuel.

This revision does not impact the f unc ti on o f the reactor core or its responto to accidents. There is no impact on reactor structure, core flow, or procedures. Removat of the startup source holders precludes the potential for holder feltures and the resultant dispersat of tecken parts within the reactor vessel.

SAFETY EVALUATIONS PROC [DURLS, TESTS AND L AFER!MENTS Page 60 The following Technical Sr.ecification M endments were incorporated into Revision 5 of the UFBAR. The NRC safety evaluation (which is t>a s e d on the Detroit Edison evaluation supporting the change) that accompanies each amendment provides the basis and justification for the UF6AR revision.

T.S. h endment Description _

UF6AR Sec t l on/T atd o 66 Use of halfnium 4.1 in Control Rt,ds 4.2 4.5

\

67 Rod Sequence Control 1.2 System (RSCS) Removal 3.1 7.1 7.6 U.15.4 Table 7.6-12 Table 7.6-13 69 Maaimum Entended 4.4 Operating Domain (MEOD) 6.3 7.6 7.7 0.15.17 Table 7.6-9 Table 7.6-10 Tables B 15.17-1 thru D.16.17-8 74 Low Pressure Cootant 6.3 Injection (LPCI) Response 15.6 Times Table 6.3-06 Table 6.3-07 77 Pressure / Temperature 3.1 Curves 4.3 5.2 A1.99 Table 4,3-3 Table 4.3-4 Table 5.2-7 Table 5.2-8 Table 5.2-0 Table 5,2-10 E ND OF SAFETY LYALUAT ION SILWHY HIPORT 6-

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